[Federal Register Volume 68, Number 62 (Tuesday, April 1, 2003)]
[Notices]
[Pages 15756-15771]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-7489]
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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 March 7, 2003 through March 20, 2003. The
last biweekly notice was published on March 18, 2003 (68 FR 12946).
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
[[Page 15757]]
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By May 1, 2003, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714,\1\ which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text to 10 CFR 2.714(d), please see 67 FR 20884; April 29,
2002.
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As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected].
[[Page 15758]]
A copy of the request for hearing and petition for leave to intervene
should also be sent to the Office of the General Counsel, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and because of
continuing disruptions in delivery of mail to United States Government
offices, it is requested that copies be transmitted either by means of
facsimile transmission to 301-415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: May 10, 2002, as supplemented March 12,
2003.
Description of amendment request: Carolina Power & Light Company
(the licensee) is proposing changes to Appendix A, Technical
Specifications (TS), and appendix B, Additional Conditions, of Facility
Operating License No. DPR-23 for the H. B. Robinson Steam Electric
Plant, Unit No. 2 (HBRSEP2). These changes will revise the licensing
basis for HBRSEP2 to implement the Alternative Source Term (AST)
described in Regulatory Guide 1.183, ``Alternative Radiological Source
Terms for Evaluating Design Basis Accidents at Nuclear Power
Reactors.'' Implementation of the AST will allow for removal of the
cycle operating length restriction from appendix B, Additional
Conditions, of the Operating License, as the AST radiological
consequence analyses support operation for an entire cycle at the
increased power level approved in License Amendment No. 196. The AST is
used by the licensee in evaluating the radiological consequences of the
following Updated Final Safety Analysis Report Chapter 15 accidents:
[sbull] Main Steam Line Break,
[sbull] Reactor Coolant Pump Shaft Seizure,
[sbull] Single Rod Control Cluster Assembly Withdrawal,
[sbull] Steam Generator Tube Rupture,
[sbull] Large Break Loss-of-Coolant Accident, and
[sbull] Waste Gas Decay Tank Rupture.
In addition, revised atmospheric dispersion factors for onsite and
offsite dose consequences have been calculated and incorporated in the
reanalysis of these events.
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:
An evaluation of the proposed change has been performed in
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards
considerations using the standards in 10 CFR 50.92(c). A discussion of
these standards as they relate to this amendment request follows:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
Implementation of the Alternative Source Term does not affect
the design or operation of HBRSEP, Unit No. 2. Rather, once the
occurrence of an accident has been postulated, the new source term
is an input to evaluate the consequences of the postulated accident.
The implementation of the Alternative Source Term has been evaluated
in revisions to limiting design basis accidents at HBRSEP, Unit No.
2. Based on the results of these analyses, it has been demonstrated
that, with the requested changes to the Technical Specifications,
the dose consequences of these limiting events are within the
regulatory guidance provided by the NRC. This guidance is presented
in 10 CFR 50.67 and Regulatory Guide 1.183. The proposed Technical
Specifications changes result in more restrictive requirements and
support the revisions to the limiting design basis accident
analyses.
Therefore, this 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 Previously Evaluated.
The proposed changes do not affect plant structures, systems or
components. The Alternative Source Term and those plant systems
affected by implementing the proposed changes do not initiate design
basis accidents.
Thus, this 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 the Margin of Safety.
The proposed changes are associated with the implementation of a
new licensing basis for HBRSEP, Unit No. 2. The new licensing basis
implements an Alternative Source Term in accordance with 10 CFR
50.67 and the associated Regulatory Guide 1.183. The results of the
revised limiting design basis analyses are subject to revised
acceptance criteria. The analyses have been performed using
conservative methodologies in accordance with the regulatory
guidance. The dose consequences of the limiting design basis events
are within the acceptance criteria found in the regulatory guidance
associated with Alternative Source Terms.
The proposed changes continue to ensure that doses at the
exclusion area and low population zone boundaries, as well as the
control room, are within the corresponding regulatory limits.
Specifically, the margin of safety for the radiological consequences
of these accidents is considered to be that provided by meeting the
applicable regulatory limits, which are conservatively set below the
10 CFR 50.67 limits. With respect to control room personnel doses,
the margin of safety (the difference between the 10 CFR 50.67 limits
and the regulatory limits defined by 10 CFR 50, Appendix A, [General
Design] Criterion 19 (GDC-19)) continues to be satisfied.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based on the above discussion, Progress Energy Carolinas, Inc.,
also known as Carolina Power and Light Company, has determined that
the requested change does not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602-1551.
NRC Section Chief: Allen G. Howe.
[[Page 15759]]
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: February 17, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specification Surveillance Requirement 3.10.1.9 to
require that the Standby Shutdown Facility (SSF) diesel generator (DG)
be loaded to at least 3280 kilowatts during the surveillance. The
current requirement is that the SSF DG be loaded to at least 3000
kilowatts during the surveillance. The change supports resolution of an
Oconee design basis issue associated with SSF pressurizer heater
capacity.
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:
Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the
determination that this amendment request involves a No Significant
Hazards Consideration by applying the standards established by the
NRC in 10 CFR 50.92. This ensures 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:
This change revises the loading of the Standby Shutdown Facility
(SSF) Diesel Generators (DG) to = 3280 kW. The design
rating of the DG is currently 3500 kW. Since the proposed loading is
within the design rating already evaluated, this proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any kind of accident previously evaluated:
As stated above, the proposed revision revises the DG loading to
an analytical value that is within the equipment's design limit.
Applicable load and support system calculations have been revised
and results have shown that the increase does not adversely affect
the ability of the SSF diesel generator or SSF to perform its
intended safety function. Additionally, this change is bounded by
all of the existing accidents and does not create the possibility of
a new or different kind of accident from any kind of accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change does not adversely affect any plant safety
limits, set points, or design parameters. The change also does not
adversely affect the fuel, fuel cladding, Reactor Coolant System, or
containment integrity. Therefore, the proposed change does not
involve a 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: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: January 31, 2003.
Description of amendment request: The proposed amendments would
revise Appendix A, Technical Specifications (TS) 3.4.11, ``RCS Pressure
and Temperature (P/T) Limits,'' to incorporate revised P/T curves. The
revised P/T curves are based on calculations performed in accordance
with General Electric (GE) Topical Report NEDC-32983P, ``General
Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux
Evaluation.'' The NEDC-32983P methodology is consistent with the
guidance contained in Regulatory Guide (RG) 1.190, ``Calculational and
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,''
dated March 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes request for LaSalle County Station, Units 1
and 2, that the pressure and temperature (P/T) limit curves in TS
3.4.11, ``RCS Pressure and Temperature (P/T) Limits,'' and
Surveillance Requirement (SR) 3.4.11.1 and SR 3.4.11.2 be revised.
The revised curves were developed using the methodology of GE
Topical Report NEDC-32983P, ``General Electric Methodology for
Reactor Pressure Vessel Fast Neutron Flux Evaluation.'' NEDC-32983P
methodology has been previously approved by the NRC for use by
licensees. The P/T limits are prescribed during normal operation to
avoid encountering pressure, temperature, and temperature rate of
change conditions that might cause undetected flaws to propagate and
cause nonductile failure of the reactor coolant pressure boundary, a
condition that is unanalyzed. Thus, the proposed changes do not have
any affect on the probability of an accident previously evaluated.
The P/T curves are used as operational limits during heatup or
cooldown maneuvering, when pressure and temperature indications are
monitored and compared to the applicable curve to determine that
operation is within the allowable region. The P/T curves provide
assurance that station operation is consistent with previously
evaluated accidents. Thus, the radiological consequences of any
accident previously evaluated are not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not change the response of plant
equipment to transient conditions. The proposed changes do not
introduce any new equipment, modes of system operation or failure
mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes adopt P/T curves that have been developed
using the methodology of GE Topical Report NEDC-32983P. The NEDC-
32983P methodology is consistent with the guidance contained in RG
1.190, ``Calculational and Dosimetry Methods for Determining
Pressure Vessel Neutron Fluence,'' dated March 2001. In a letter
dated September 14, 2001, the NRC approved NEDC-32983P for use by
licensees.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Based upon the above, EGC concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief : Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: February 27, 2003.
[[Page 15760]]
Description of amendment request: The proposed amendments revise
the Technical Specifications to reflect a one-time deferral of the
primary containment Type A leak rate test to no later than July 22,
2009, for Unit 1 and no later than May 16, 2008, for Unit 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will revise Quad Cities Nuclear Power
Station (QCNPS), Units 1 and 2, Technical Specification (TS) 5.5.12,
``Primary Containment Leakage Rate Testing Program,'' to reflect a
one-time deferral of the primary containment Type A test to no later
than July 22, 2009, for Unit 1, and no later than May 16, 2008, for
Unit 2. The current Type A test interval of 10 years, based on past
performance, would be extended on a one-time basis to 15 years from
the last Type A test.
The function of the primary containment is to isolate and
contain fission products released from the reactor coolant system
(RCS) following a design basis loss of coolant accident (LOCA) and
to confine the postulated release of radioactive material to within
limits. The test interval associated with Type A testing is not a
precursor of any accident previously evaluated. Therefore, extending
this test interval on a one-time basis from 10 years to 15 years
does not result in an increase in the probability of occurrence of
an accident. The successful performance history of Type A testing
provides assurance that the QCNPS primary containments will not
exceed allowable leakage rate values specified in the TS and will
continue to perform their design function following an accident. The
risk assessment of the proposed change has concluded that there is
an insignificant increase in total population dose rate and an
insignificant increase in the conditional containment failure
probability.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change for a one-time extension of the Type A tests
for QCNPS, Units 1 and 2, will not affect the control parameters
governing unit operation or the response of plant equipment to
transient and accident conditions. The proposed change does not
introduce any new equipment or modes of system operation. No
installed equipment will be operated in a new or different manner.
As such, no new failure mechanisms are introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
QCNPS, Units 1 and 2, are General Electric BWR/3 [boiling water
reactor class 3] plants with Mark I primary containments. The Mark I
primary containment consists of a drywell, which encloses the
reactor vessel, reactor coolant recirculation system, and branch
lines of the RCS; a toroidal-shaped pressure suppression chamber
containing a large volume of water; and a vent system connecting the
drywell to the water space of the suppression chamber. The primary
containment is penetrated by access, piping, and electrical
penetrations.
The integrity of the primary containment penetrations and
isolation valves is verified through Type B and Type C local leak
rate tests (LLRTs) and the overall leak tight integrity of the
primary containment is verified by a Type A integrated leak rate
test (ILRT) as required by 10 CFR 50, Appendix J, ``Primary Reactor
Containment Leakage Testing for Water-Cooled Power Reactors.'' These
tests are performed to verify the essentially leak tight
characteristics of the primary containment at the design basis
accident pressure. The proposed change for a one-time extension of
the Type A tests do not affect the method for Type A, B, or C
testing, or the test acceptance criteria. In addition, based on
previous Type A testing results, EGC [Exelon Generation Company,
LLC] does not expect additional degradation, during the extended
period between Type A tests, which would result in a significant
reduction in a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: February 27, 2003.
Description of amendment request: The proposed amendments add a
surveillance requirement to perform a quarterly trip unit calibration
of the reactor protection system scram discharge volume water level--
high differential pressure switches.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specifications (TS) change adds a trip
unit calibration surveillance requirement (SR) for the analog trip
units associated with the Scram Discharge Volume (SDV) Water Level--
High Trip Function for the Reactor Protection System (RPS)
Instrumentation. Specifically, SR 3.3.1.1.11 is added to Function
7.b of TS Table 3.3.1.1-1, ``Reactor Protection System
Instrumentation.'' In addition, the proposed change revises Function
7.a of TS Table 3.3.1.1-1 to delete a reference to thermal switches,
applicable to Unit 1 through cycle 17. The change to Function 7.a is
editorial, since Unit 1 SDV level instrumentation has been upgraded
to replace Fluid Components International thermal switches with
Magnetrol float switches.
TS requirements that govern operability or routine testing of
plant instruments are not assumed to be initiators of any analyzed
event because these instruments are intended to prevent, detect, or
mitigate accidents. Therefore, these proposed changes will not
involve an increase in the probability of an accident previously
evaluated. Additionally, these proposed changes do not increase the
consequences of an accident previously evaluated because the
proposed changes do not adversely impact structures, systems, or
components. The proposed changes establish requirements that ensure
components are operable when necessary for the prevention or
mitigation of accidents or transients. Furthermore, there will be no
change in the types or significant increase in the amounts of any
effluents released offsite.
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
There is no change being made to the parameters within which
Quad Cities Nuclear Power Station (QCNPS) is operated. The proposed
changes do not adversely impact the manner in which the SDV Water
Level--High RPS instrumentation will operate under normal and
abnormal operating conditions. The proposed changes will not alter
the function demands on credited equipment. No alteration in the
procedures, which ensure QCNPS remains within analyzed limits, is
proposed, and no change is being made to procedures relied upon to
respond to an off-normal event. Therefore, these proposed changes
provide an equivalent level of safety and will not create the
possibility of a new or different
[[Page 15761]]
kind of accident from any accident previously evaluated. The changes
in methods governing normal plant operation are consistent with the
current safety analysis assumptions. Therefore, these proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety[?]
Response: No.
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of setpoints to initiate alarms
and actions. The proposed changes do not affect the probability of
failure or availability of the affected instrumentation, and the
proposed changes do not revise any allowable values for RPS
functions. Therefore, it is concluded that the proposed changes do
not result in a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: January 14, 2003.
Description of amendment request: This license amendment request
proposes a change to Technical Specifications (TSs) 5.1.1, 5.4.1, and
5.5.1 that would replace the requirement for the plant manager to
approve administrative procedures and the Offsite Dose Calculation
Manual. The plant manager approval signature would be replaced with the
signature of a procedurally authorized individual who would be a more
appropriate authority for approval of the activity.
Basis for proposed no significant hazards consideration
determination: As required by Section 50.91(a) of Title 10 of the Code
of Federal Regulations (10 CFR), 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 to replace the plant manager's approval with
the approval by an authorized individual is consistent with the
requirements of Regulatory Guide 1.33 and American National
Standards Institute (ANSI) N18.7-1976/American Nuclear Society (ANS)
3.2. The authorized individuals are management and supervisory
personnel who satisfy the requirements of ANSI N18.1-1971. Use of
ANSI N18.1-1971 is consistent with the requirements of the existing
TS and Updated Safety Analysis Report (USAR). The change is
administrative and does not impact or otherwise affect the physical
plant.
The proposed change to the License Condition to delete the
reporting time frame eliminates duplication of a requirement that is
already an integral part of 10 CFR 50.73 which is referenced in the
License Condition. The proposed change is administrative and does
not impact or otherwise affect the physical plant.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated. The
proposed administrative changes do not involve any physical
modifications to the facility nor add new equipment. The methods of
plant operation have not been altered. 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 changes will not involve a significant reduction
in the margin of safety.
The proposed changes are administrative in nature and have no
direct impact upon any plant safety analyses. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant (PNPP), Unit 1, Lake County, Ohio
Date of amendment request: January 30, 2003.
Description of amendment request: This license amendment request
would modify the existing minimum critical power ratio (MCPR) safety
limit contained in Technical Specification (TS) 2.1.1.2. Specifically,
the change modifies the MCPR safety limit values, as calculated by
Global Nuclear Fuel (GNF), by decreasing the limit for two
recirculation loop operation from 1.10 to 1.07, and decreasing the
limit for single recirculation loop operation from 1.11 to 1.08. The
change resulted from the core reload analysis performed for the Perry
Nuclear Power Plant (PNPP) fuel cycle 10.
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.
PPNP Updated Safety Analysis Report (USAR) Section 4.2, ``Fuel
System Design,'' states the PNPP fuel system design bases are
provided in the General Electric Topical Report, NEDE-24011-P-A,
``General Electric Standard Application for Reactor Fuel (GESTAR
II).'' The MCPR Safety Limit is one of the limits used to protect
the fuel in accordance with the design basis. The MCPR Safety Limit
establishes a margin to the onset of transition boiling. The basis
of the MCPR Safety Limit remains the same, ensuring that greater
than 99.9% of all fuel rods in the core avoid transition boiling.
The methodology used to determine the MCPR Safety Limit values is
contained within GESTAR II and is NRC approved. The change does not
result in any physical plant modifications or physically affect any
plant components. As a result, there is no increase in the
probability of occurrence of a previously analyzed accident.
The fundamental sequences of accidents and transients have not
been altered. The Safety Limit MCPR is established to avoid fuel
damage in response to anticipated operational occurrences.
Compliance with a MCPR Safety Limit greater than or equal to the
calculated value will ensure that less than 0.1% of the fuel rods
will experience boiling transition. This in turn ensures fuel damage
does not occur following transitions due to excessive thermal
stresses on the fuel cladding. The MCPR Operating Limits are set
higher (i.e., more conservative) than the Safety Limit such that
potentially limiting plant transients prevent the MCPR from
decreasing below the MCPR Safety Limit during the transient.
Therefore, there is no impact on any limiting USAR Appendix 15B
transients. The radiological consequences remain the same as
previously stated in the USAR. Therefore, the consequences of an
accident do not increase over previous evaluations in the USAR.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The MCPR Safety Limit basis is preserved, which is to ensure
that transition boiling does not occur in at least 99.9% of the fuel
rods in the core as a result of the postulated limiting transient.
The values are calculated in accordance with GESTAR II. The GESTAR
II analyses have been accepted by the NRC.
[[Page 15762]]
The MCPR Safety Limit is one of the limits established to ensure the
fuel is protected in accordance with the design basis. The function,
location, operation, and handling of the fuel remain unchanged. No
changes in the design of the plant or the method of operating the
plant are associated with these revised safety limit values.
Therefore, no new or different kind of accident from any previously
evaluated is created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This change revises the PNPP MCPR Safety Limit values. The new
MCPR Safety Limit values reflect changes due to the Cycle 10 core
reload, but do not alter the design or function of any plant system,
including the fuel. The new MCPR Safety Limit values were calculated
using NRC-approved methods described in GESTAR II. The proposed MCPR
Safety Limit values continue to satisfy the fuel design safety
criteria which ensures that transition boiling does not occur in at
least 99.9% of the fuel rods in the core as a result of the
postulated limiting transient. Therefore, the proposed values for
the MCPR Safety Limit 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of amendment request: December 10, 2002.
Description of amendment request: The proposed amendment would
revise the Unit 2 reactor coolant system (RCS) pressure-temperature
curves in Technical Specification (TS) Figures 3.4-2 and 3.4-3 and
associated TS Bases. The revised curves will bound operation of the
unit for the remainder of its current license duration and bound
operation with planned license amendments to increase the power level
at which the unit is allowed to operate. In support of this proposed
amendment, Indiana Michigan Power (I&M) has submitted a request, in
accordance with 10 CFR 50.60, ``Acceptance Criteria for Fracture
Prevention Measures for Lightwater Nuclear Power Reactors for Normal
Operation,'' for exemption from requirements in 10 CFR Part 50,
Appendix G, ``Fracture Toughness Requirements.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
Probability of Occurrence of an Accident Previously Evaluated
The proposed change will revise the RCS pressure-temperature
curves to reflect new limiting reactor vessel materials, to bound
operation of the reactor up to 3600 MWt for the current fuel cycle
and beyond, to reflect new fluence analysis methodology, to reflect
the use of ASME [American Society of Mechanical Engineers] Code Case
N-641, to include boltup limits, and to no longer include instrument
uncertainty margins.
The proposed change will not result in physical changes to
structures, systems, or components (SSCs), or changes to event
initiators or precursors. The proposed change will not affect the
ability of personnel to control RCS pressure at low temperatures
and, thereby, ensure the integrity of the reactor coolant pressure
boundary. Use of Code Case N-641 in developing the proposed revision
to the RCS pressure-temperature curves is in accordance with
methodologies accepted by the ASME. These methodologies provide
assurance that the reactor vessel will withstand the effects of
normal cyclic loads due to temperature and pressure changes, and
provide an acceptable level of protection against brittle failure.
Additionally, the proposed changes will not impact the design or
operation of plant systems such that previously analyzed SSCs will
be more likely to fail. The initiating conditions and assumptions
for accidents described in the UFSAR [updated final safety analysis
report] will remain as previously analyzed. Therefore, the proposed
changes will not involve a significant increase in the probability
of an accident previously evaluated.
Consequences of an Accident Previously Evaluated
The proposed change does not reduce the ability of any SSC to
limit the radiological consequences of accidents described in the
UFSAR. The proposed change will not alter any assumptions made in
the analysis of radiological consequences of previously evaluated
accidents, nor does it affect the ability to mitigate these
consequences. No new or different radiological source terms will be
generated as a result of the proposed change. Therefore, the
proposed changes do not involve a significant increase in the
consequences of an accident previously evaluated.
The format changes will improve the appearance of the affected
pages but will not affect any requirements. In summary, the
probability of occurrence and the consequences of an accident
previously evaluated will not be significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not result in physical changes to SSCs.
The proposed change will not involve the addition or modification of
plant equipment (no new or different type of equipment will be
installed) nor will it alter the design of any plant systems. The
proposed change solely involves RCS pressure-temperature limits. The
types of potential accidents associated with these limits have been
previously identified and evaluated. No new accident scenarios,
accident or transient initiators or precursors, failure mechanisms,
or single failures will be introduced as a result of the proposed
changes. No new or different modes of failure will be created. The
format changes will improve the appearance of the affected pages but
will not affect any requirements. Therefore, the proposed change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed RCS pressure-temperature curves will continue to
provide adequate margins of protection for the reactor coolant
pressure boundary. The proposed changes have been determined,
through supporting analyses, to be in accordance with the
methodologies and criteria set forth in the applicable regulations,
or in accordance with technically adequate alternatives. Compliance
with these methodologies provides adequate margins of safety and
ensures that the reactor coolant pressure boundary will withstand
the effects of normal cyclic loads due to temperature and pressure
changes as well as the loads associated with postulated faulted
events as described in the UFSAR. The format changes will improve
the appearance of the affected pages but will not affect any
requirements. Therefore, the proposed change will not significantly
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 requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS),
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: February 25, 2003.
Description of amendment request: The proposed Technical
Specification
[[Page 15763]]
(TS) changes will add an allowed outage time (AOT) for Engineered
Safety Features Actuation System (ESFAS) Instrumentation channels to be
out of service in a bypassed state.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The addition of an ACTION STATEMENT and the addition of an AOT
(and its associated actions if not met) for a TS action statement
are neither an accident initiator nor precursor. The ESFAS actuates
in response to an accident and has a mitigating function. Increasing
the TS requirements for specific TS instrument loops provides
additional assurance that the channels will be capable of performing
their design function in the event of a DBA [design-basis accident].
The ability of the operations staff to respond to an evaluated
accident or plant transient will not be hampered. This change
provides conservative requirements to assure that the design basis
of the plant is maintained.
Addition of conservative changes to the Engineered Safety
Feature Actuation System Instrumentation does not contribute to the
initiation of any accident evaluated in the FSAR [Final Safety
Analysis Report]. Supporting factors are as follows:
[sbull] The changes provide consistency between Tables 3.3-2,
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the
functional units in those tables. These changes are conservative and
consistent with the Standard Technical Specifications, NUREG-1431,
Rev. 2. There are no deletions from the Technical Specifications
made by these changes, nor relaxation in any applicability, action,
or surveillance requirements.
[sbull] Overall plant performance and operation is not altered
by the proposed changes. There are to be no plant hardware changes
as a result of this proposed change and only minimal procedural
changes.
Therefore, since the Engineered Safety Feature Actuation System
Instrumentation are treated more conservatively, the probability of
occurrence or consequences of an accident evaluated in the VCSNS
FSAR will be no greater than the original design basis of the plant.
Therefore, the change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes provide consistency between Tables 3.3-2,
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the
functional units in those tables. Additionally, the addition of an
ACTION STATEMENT and an AOT with conservative requirements are
intended to assure that the plant is in a safe configuration and can
meet accident analyses assumptions. These changes are conservative
and consistent with the Improved Technical Specifications, NUREG-
1431, Rev. 2. No new accident initiator mechanisms are introduced
since:
[sbull] No physical changes to the Engineered Safety Feature
Actuation System Instrumentation are made.
[sbull] No deletions from the Technical Specifications are made.
[sbull] No relaxations in any applicability, action, or
surveillance requirements are made.
Since the safety and design requirements continue to be met and
the integrity of the reactor coolant system pressure boundary is not
challenged, no new accident scenarios have been created. Therefore,
the types of accidents defined in the FSAR continue to represent the
credible spectrum of events to be analyzed, which determine safe
plant operation.
3. Does this change involve a significant reduction in margin of
safety?
The proposed change requires that an instrument channel for an
Engineered Safety Feature remain operable or be restored to
operability within a reasonable time period, otherwise a controlled
shutdown is required. This conforms to the safety analysis where the
plant and its systems, structures and components must be capable of
performing the safety function while a DBA is occurring, in the
presence of a worst case single failure.
This is not a reduction in a margin of safety, since it restores
the margin that was designed into the plant.
The proposed changes provide consistency between Tables 3.3-2,
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the
functional units in those tables. These changes are conservative and
consistent with the Standard Technical Specifications, NUREG-0452,
Rev. 5. The proposed changes impose more restrictive operating
limitations, and their use provides increased assurance that the
Engineered Safety Feature Actuation System Instrumentation remains
operable. Since the changes are conservative additions, it is
concluded that the changes do not involve a significant reduction in
the margin of safety. This is not a reduction in a margin of safety,
since it restores the margin that was designed into the plant.
Pursuant to 10 CFR 50.91, the preceding analyses provides a
determination that the proposed Technical Specifications change
poses no significant hazard as delineated by 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas G. Eppink, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: John A. Nakoski.
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50-
296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone
County, Alabama
Date of amendment request: February 13, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 4.2.1, Fuel Assemblies, to modify
the fuel design description to encompass Framatome Advanced Nuclear
Power (FANP) fuel assemblies and also to modifiy TS 4.3, Fuel Storage,
to remove nomenclature specific to Global Nuclear Fuels analysis
methods. The proposed TS changes are needed to allow the receipt and
storage of Framatome fuel assemblies.
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 in accordance with the three standards set forth in 10
CFR 50.92(c), which are presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed amendment revises TS 4.2.1, Fuel Assemblies, to
modify the fuel design description to accommodate FANP fuel designs.
The change to TS 4.2.1 is administrative and simply adds descriptive
text to reflect that FANP fuel assemblies have a water channel.
To make the fuel storage TS compatible with the storage of GNF
[Global Nuclear Fuels] and FANP fuel, the proposed amendment also
modifies TS 4.3, Fuel Storage, to delete criteria specific to GNF
fuel storage criticality analysis methods. BFN criticality analysis
and storage requirements continue to be adequately described in the
Updated Final Safety Analysis Report (UFSAR) and in existing TS
4.3.1.1.b, TS 4.3.1.1.c, TS 4.3.1.2.b, 4.3.1.2.c, and 4.3.1.2.d.
Hence, the proposed elimination of the GNF-specific criteria in TS
4.3 does not affect BFN design basis requirements associated with
ensuring adequate criticality margins are maintained for fuel
storage.
The requested TS changes do not involve any plant modifications
or operational changes that could affect system reliability,
performance, or the possibility of operator error. The requested
changes do not affect any postulated accident precursors, do not
affect accident mitigation systems, and do not introduce any new
accident initiation methods. Therefore, the proposed TS change does
not involve an increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed changes to TS do not affect the performance of
any BFN structure,
[[Page 15764]]
system, or component credited with mitigating any accident
previously evaluated. Fuel storage criticality analyses will
continue to be performed in accordance with established UFSAR
commitments that are independent are fuel vendor specific methods.
The TS changes do not introduce new modes of operation or involve
plant modifications.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed amendment modifies TS 4.3, Fuel Storage, to
remove nomenclature specific to GNF criticality analysis methods.
Fuel storage criticality analyses will continue to be performed in
accordance with UFSAR commitments and the remaining TS commitments
in accordance with FANP accepted methods, which specify appropriate
criteria and conservatisms. Therefore, the proposed TS change does
not involve a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant (WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: December 19, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Chapter 5.0, ``Administrative
Controls,'' to incorporate three approved TS Task Force (TSTF) changes:
TSTF-258, Revision 4; TSTF-299, Revision 0; and TSTF-308, Revision 1.
These changes have been incorporated in Revision 2 of NUREG 1431,
``Standard Technical Specifications Westinghouse Plants.''
In addition, the amendment proposes two editorial changes. These
changes either update personnel titles with the titles currently used
at WBN and TVA's other nuclear units or clarify required staffing
levels.
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 in accordance with the three standards set forth in 10
CFR 50.92(c), which are presented below:
1. Does the proposed change involve a significant increase in
the probability of consequences of an accident previously evaluated?
No. The proposed changes affect only administrative requirements
or programs. As indicated below, the justification for five of the
changes (Parts 2 through 4 of Change Number 2 and Change Numbers 3,
5 [only Parts 1 and 2 of Change 5], 6, and 7) is based on the
existence of a regulation or other regulatory document which
controls the administrative requirements. For these changes, the
proposed amendment modifies the administrative TS to make it
consistent with the current regulations or NRC guidance document.
Two changes (Change Number 1 and Part 1 of Change Number 2) are
strictly editorial. In addition, two changes (Change Number 4 and
Part 3 of Change Number 5) add a requirement to make the program
consistent with the criteria for Surveillance Requirements in the
Improved Standard Technical Specifications (ISTS). Based on the
preceding information, the proposed amendment does not involve
technical changes to the configuration or operation of the plant
there is not a significant increase in the probability or
consequences of an accident previously evaluated:
------------------------------------------------------------------------
Administrative Justification for the
Change No. section affected change
------------------------------------------------------------------------
1............................. 5.1, Editorial update of
``Responsibility staff titles.
,'' Section
5.1.2.
2............................. 5.2.2, ``Unit Part 1 of Change
Staff''. number 2--Editorial
clarification of the
number of non-
licensed operators
required for the
operation of WBN
Unit 1. Parts 2
through 4 of Change
Number 2--The
existing
administrative
requirements are
revised to align the
requirements with 10
CFR 50.54.
3............................. 5.3, ``Unit Staff Adds TS 5.3.2 which
Qualifications,' clarifies the
' Section 5.3.2. ``Operator'' and
``Senior Operator''
definitions in 10
CFR 55.4 and ties
these positions to
the requirements of
10 CFR 50.54.
4............................. 5.7.2.4, WBN TS 5.7.2.4 serves
``Primary the same function as
Coolant Sources a Surveillance
Outside Requirement (SR).
Containment. The proposed change
structures TS
5.7.2.4 so that it
is consistent with
other ISTS SRs and
the frequency
extension allowed by
SR 3.0.2.
5............................. 5.7.2.7, The intent of the
``Radioactive revisions to this TS
Effluent are to: 1) eliminate
Controls possible confusion
Program''. or improper
implementation of
the requirements of
10 CFR 20; 2)
clarifies the
wording to not
require dose
projections for a
calendar quarter and
a calendar year
every 31 days; 3)
structures the TS so
that it is
consistent with
other ISTS SRs.
6............................. 5.9.4, ``Monthly The proposed change
Operating makes the TS
Reports''. reporting
requirements
consistent with the
reporting
requirements in
Generic Letter 97-
02.
7............................. 5.11, ``High The proposed revision
Radiation Area''. updates the TS to be
consistent with 10
CFR 20.1601(c) and
updates the
acceptable alternate
controls to those
given in 10 CFR
20.1601.
------------------------------------------------------------------------
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. As indicated above, the proposed changes do not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) or changes in methods controlling
normal plant operation. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed changes will not reduce the margin of safety
because they have no effect on assumptions made in WBN's safety
analysis or the configuration of plant equipment important to
safety. Additionally, several of the proposed revisions adjust the
administrative requirements to be consistent with existing
regulations or NRC guidance documents and therefore, will not
adversely impact plant safety. The balance of the proposed changes
are editorial updates or adjust a program to be consistent with the
ISTS.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
[[Page 15765]]
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority (TVA), Docket No. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: February 14, 2003.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) for Watts Bar Nuclear Plant
(WBN), Unit 1. The proposed TS change would allow WBN Unit 1 to be
refueled and operated using the Westinghouse 17x17 Robust Fuel
Assembly-2 (RFA-2) design commencing with Cycle 6 in September 2003.
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. The Loss of Coolant Accident (LOCA) and non-LOCA transients
and accidents which are potentially affected by the parameters and
assumptions associated with the use of RFA-2 (including the effects
of Tritium Producing Burnable Absorber Rods, TPBARs) have been
evaluated/analyzed and all design standards and applicable safety
criteria are met. The consideration of these changes does not result
in a situation where the design material, and construction standards
that were applicable prior to the change are altered. Therefore, the
changes occurring with the use of RFA-2 will not result in any
additional challenges to plant equipment that could increase the
probability of any previously evaluated accident.
The changes associated with the use of RFA-2 do not affect plant
systems such that their function in the control of radiological
consequences is adversely affected. TVA's evaluation documents that
the design standards and applicable safety criteria limits continue
to be met and, therefore, fission barrier integrity is not
challenged. The fuel rod design (the first fission product barrier)
is not changed. Compared to the current grid design on the resident
fuel, the RFA-2 grid design provides improved resistance to fuel rod
fretting. The RFA-2 fuel changes have been shown not to adversely
affect the response of the plant to postulated accident scenarios.
These changes will therefore not affect the mitigation of the
radiological consequences of any accident described in the Final
Safety Analysis Report (FSAR).
Therefore, since the actual plant configuration, performance of
systems, and initiating event mechanisms are not being changed as a
result of this evaluation, TVA has concluded that the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated.
No. The possibility for a new or different type of accident from
any accident previously evaluated is not created since the changes
associated with the use of RFA-2 do not result in a change to the
design basis of any plant component or system. The evaluation of the
effects of the use of RFA-2 shows that all design standards and
applicable safety criteria limits are met. Specifically, the results
of the evaluations/analyses lead to the following conclusions:
1. The RFA-2 fuel design for Watts Bar Unit 1 is mechanically
compatible with the current fuel assemblies, core components, the
control rods and the reactor internals interfaces.
2. The structural integrity of the RFA-2 fuel design has been
evaluated for seismic/LOCA loadings for Watts Bar Unit 1. Evaluation
of the RFA-2 fuel assembly component stresses and grid impact forces
due to postulated faulted condition accidents verified that the fuel
assembly design is structurally acceptable.
3. The changes to the nuclear characteristics due to the
transition to the RFA-2 fuel assembly design will be within the
range normally seen from cycle to cycle due to fuel management.
4. The RFA-2 fuel assembly design is hydraulically compatible
with the current fuel assemblies.
5. The core design and safety analyses documented in this report
demonstrate the capability of the core to operate safely at the
rated Watts Bar Unit 1 design thermal power with either a mixed core
of RFA-2 fuel and the current fuel product or with a full core of
RFA-2 fuel.
6. TVA's amendment request establishes a reference upon which to
base Westinghouse reload safety evaluations for future reloads with
the RFA-2 fuel assembly design.
7. Reload core designs with either a mixed core of RFA-2 fuel
and the current fuel product or with a full core of RFA-2 fuel are
compatible with the planned introduction of Tritium-Producing
Burnable Absorber Rods (TPBARs) into Watts Bar Unit 1.
These changes therefore do not cause the initiation of any
accident nor create any new failure mechanisms. All equipment
important to safety will operate as designed. Component integrity is
not challenged. The changes do not result in any event previously
deemed incredible being made credible. The use of RFA-2 is not
expected to result in more adverse conditions and is not expected to
result in any increase in the challenges to safety systems.
Therefore, TVA concludes that this proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety.
No. The margin of safety is maintained by assuring compliance
with acceptance limits reviewed and approved by the NRC. All of the
appropriate acceptance criteria for the various analyses and
evaluations have been met, therefore, there has not been a reduction
in any margin of safety.
Therefore, TVA concludes that 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide
[[Page 15766]]
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: April 10, 2002.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to relocate emergency diesel generator maintenance
inspection requirements from Section 4.7 to the Updated Final Safety
Analysis Report.
Date of Issuance: March 7, 2003.
Effective date: March 7, 2003 shall be implemented within 30 days
of issuance, except the relocation of the emergency diesel generator
maintenance requirements of Technical Specification 4.7, which shall be
incorporated into the Updated Final Safety Analysis Report in
accordance with the schedule specified by 10 CFR 50.71.
Amendment No.: 236.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 28, 2002 (67 FR
36926).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: August 30, 2002, as supplemented
November 21 and December 16, 2002, and January 23, 2003.
Brief description of amendment: This amendment revises the
Technical Specifications by eliminating the requirements to perform
response time testing for several reactor protection system and
engineered safety feature functions in conformance with previously
approved topical reports.
Date of issuance: March 7, 2003.
Effective date: March 7, 2003.
Amendment No.: 112.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 1, 2002 (67 FR
61676).
The November 21 and December 16, 2002, and January 23, 2003,
letters provided clarifying information and did not change the initial
proposed no significant hazards consideration determination or expand
the scope of the initial application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1 (HNP), Wake and Chatham Counties,
North Carolina
Date of application for amendment: August 28, 2002.
Brief description of amendment: This amendment revises Technical
Specification (TS) 3/4.9.9, ``Containment Ventilation Isolation
System,'' to allow the same administrative controls for this TS as were
approved previously by the NRC in Amendment No. 104 to the HNP TS for
TS 3/4.9.4, ``Containment Building Penetrations,'' to provide
consistency between the two TS.
Date of issuance: March 12, 2003.
Effective date: March 12, 2003.
Amendment No.: 113.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61676).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear
Station, Unit 2, York County, South Carolina
Date of application for amendments: October 10, 2002, as
supplemented by letters dated February 7 and February 26, 2003.
Brief description of amendments: The amendment authorizes the
licensee to continue to use, for operational cycle 13 beginning in
March 2003, and subsequent cycles of operation, the reactor coolant
system cold leg elbow tap flow coefficients that were approved by the
NRC on an interim basis for cycle 12 in Amendment No. 186.
Date of issuance: March 19, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 199.
Facility Operating License No. NPF-52: Amendment authorizes
revision of the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: November 26, 2002 (67
FR 70765).
The supplements dated February 7 and February 26, 2003, provided
clarifying information that did not change the scope of the October 10,
2002, 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 March 19, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: May 29, 2002, as supplemented
by letters dated September 25 and November 12, 2002, and January 8 and
January 29, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications to allow a one-time change in the Appendix J,
Type A containment integrated leakage rate test interval from the
currently required 10-year interval to a test interval of 15 years.
Date of issuance: March 12, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 205/198.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45563).
The supplements dated September 25 and November 12, 2002, and
January 8 and January 29, 2003, provided clarifying information that
did not change the scope of the May 29, 2002, application or the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 12, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: May 29, 2002, as supplemented
by letters dated September 25 and
[[Page 15767]]
November 12, 2002, and January 8 and January 29, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications to allow a one-time change in the Appendix J,
Type A containment integrated leakage rate test interval from the
currently required 10-year interval to a test interval of 15 years.
Date of issuance: March 12, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 211/192.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45563).
The supplements dated September 25 and November 12, 2002, and
January 8 and January 29, 2003, provided clarifying information that
did not change the scope of the May 29, 2002, application or the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 12, 2003.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: April 24, 2002, as supplemented by
letters dated July 18, December 18 and 20, 2002, and February 19, 2003.
Brief description of amendment: The amendment reflects a full-scope
implementation of the alternative source term, as described in
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' pursuant
to 10 CFR 50.67, ``Accident source term.''
Date of issuance: March 14, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 132.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 11, 2002 (67 FR
40021).
The July 18, December 18 and 20, 2002, and February 19, 2003,
supplemental letters provided clarifying information that did not
change the scope of the original Federal Register notice or the
original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 14, 2003.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 14, 2002, as supplemented by letters
dated February 12 and 28, 2003.
Brief description of amendment: The amendment modifies the
surveillance requirements (SRs) pertaining to the testing of the
Division 3 standby emergency diesel generator (EDG). The change allows
performance of some required surveillance tests for the Division 3 EDG
during any mode of plant operation (previously allowed only in Modes 4
(Cold Shutdown) and 5 (Refueling)).
Date of issuance: March 14, 2003.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 133.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42824).
The February 12, 2003, supplemental letter provided clarifying
information and the February 28, 2003, supplemental letter withdrew the
requested change to the Note associated with SR 3.8.1.8. The
supplemental letters did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 14, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 5, 2002, as supplemented on
January 9 and March 4, 2003.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) to implement the alternate source term methodology
for the fuel-handling accident analysis. Specifically, the amendment
revises TS 3.9.3, ``Containment Penetrations,'' to: (1) Permit the
equipment closure hatch opening and the personnel airlock doors to be
capable of being closed during movement of irradiated fuel, (2) allow
use of administrative controls for unisolating containment penetrations
during movement of irradiated fuel, (3) delete the containment purge
and containment pressure relief requirements and associated
surveillances with the reactor subcritical for less than 550 hours, and
(4) eliminate the TS applicability ``during core alterations.'' In this
regard, the amendment adopts TS Task Force (TSTF) Standard TS Change
Travelers TSTF-68, ``Containment Personnel Airlock Doors Open During
Fuel Movement,'' TSTF-312, ``Administratively Control Containment
Penetrations,'' and, in part, TSTF-51, ``Revise Containment
Requirements During Handling Irradiated Fuel and Core Alterations.''
The amendment also revises the Applicability Statements for Limiting
Condition for Operation (LCO) 3.3.8 for the fuel storage building
emergency ventilation system (FSBEVS) actuation instrumentation and LCO
3.7.13 for the FSBEVS to also add the term ``recently'' before
``irradiated fuel assemblies.'' In addition, the LCO Required Action
would likewise be modified to add the term ``recently'' to now require
the suspension of movement of recently irradiated fuel in the FSB.
Date of issuance: March 17, 2003.
Effective date: March 17, 2003.
Amendment No.: 215.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45567).
The January 9 and March 4 letters provided clarifying information
that did not expand the scope of the proposed amendment or change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: August 16, 2002.
Brief description of amendment: The amendment relocates certain
Control Rod Block functions from Technical Specifications 3/4.2.C,
``Control Rod Block Actuation,'' Tables 3.2.C.1, 3.2.C-2, and 4.2.C to
the Updated Final Safety Analysis Report.
Date of issuance: March 17, 2003.
[[Page 15768]]
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 196.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68735).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 24, 2002, as supplemented by
letter dated February 4, 2003.
Brief description of amendment: The amendment revises Technical
Specifications (TSs) relating to positive reactivity additions while in
shutdown modes by clarifying TSs involving positive reactivity
additions. In addition, the borated water volume requirements in TS
3.1.2.7 is now presented in ``percent level'' units and an obsolete
reference from Surveillance Requirement 4.8.2.2 is deleted.
Date of issuance: March 7, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 185.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75874).
The February 4, 2003, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: April 2, 2001, as supplemented by
letters dated September 24, 2001, and February 27, July 31, and
December 19, 2002.
Brief description of amendment: The Refueling Water Storage Pool
(RWSP) purification system is aligned to the RWSP to maintain the
purity and clarity of the borated water contained in the pool. It is
also one of two means of makeup to the Spent Fuel Pool, with the
Condensate Storage Pool being the primary makeup source. Entergy
Operations Inc. has proposed to revise its Waterford Steam Electric
Station, Unit 3, Updated Final Safety Analysis Report (UFSAR) to allow
the manual valves (FS-423 and FS-404) that isolate the RWSP from the
RWSP purification system and provide the boundary between the
seismically qualified, safety related RWSP and the non-seismic, non-
safety related RWSP purification system to be maintained open.
Date of issuance: March 12, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 186.
Facility Operating License No. NPF-38: The amendment revised the
UFSAR.
Date of initial notice in Federal Register: May 16, 2001 (66 FR
27176).
The September 24, 2001, and February 27, July 31, and December 19,
2002, supplemental letters provided clarifying information that did not
change the scope of the original Federal Register notice or the
original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: November 27, 2002.
Brief description of amendments: These amendments delete technical
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminate the requirements to have and maintain the post accident
sampling system at the Dresden Nuclear Power Station, Units 2 and 3.
The amendments also address related changes to TS 5.5.2, ``Primary
Coolant Sources Outside Containment.''
Date of issuance: March 11, 2003.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 197/190.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2802).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 11, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania
Date of application for amendment: November 21, 2002, as
supplemented February 25, 2003.
Brief description of amendment: This amendment revised the
Technical Specifications (TSs) for the safety limit for the minimum
critical power ratio from its current value of 1.09 to 1.07 for two
recirculation-loop operations, and from 1.11 to 1.09 for single
recirculation-loop operation.
Date of issuance: March 11, 2003.
Effective date: As of the date of issuance, to be implemented prior
to startup for Cycle 8 operations, scheduled for March 2003.
Amendment No.: 127.
Facility Operating License No. NPF-85: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
802). The February 25, 2003, letter provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register on January 7, 2003 (68 FR 802).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 11, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: November 27, 2002.
Brief description of amendments: These amendments delete technical
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminate the requirements to have and maintain the post accident
sampling system at the Quad Cities Nuclear Power Station, Units 1 and
2. The amendments also address related changes to TS 5.5.2, ``Primary
Coolant Sources Outside Containment.''
Date of issuance: March 11, 2003.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 212/206.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2802)
[[Page 15769]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 11, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: March 14, 2002, as supplemented
by letters dated July 17 and September 12, 2002, and January 24, 2003.
Brief description of amendment: This amendment supplements License
Amendment No. 100, which was issued on February 24, 1999, by placing
restrictions on removing the inclined fuel transfer system (IFTS) blind
flange during Operational Modes 1, 2, and 3. The amendment includes a
time limit on the removal of the IFTS blind flange, provides a
requirement to install the upper pool IFTS gate prior to IFTS blind
flange removal, and limits the unbolted configuration of the IFTS blind
flange when it is rotated.
Date of issuance: March 7, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 123.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5675).
The supplemental information contained clarifying information that
was within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: October 30, 2002.
Brief description of amendment: This amendment deletes Technical
Specification (TS) 5.5.3, ``Post Accident Sampling System (PASS),'' and
thereby eliminates the requirements to have and maintain the PASS at
the Perry Nuclear Power Plant, Unit 1. The amendment also addresses
related changes to TS 5.5.2, ``Primary Coolant Sources Outside
Containment.''
Date of issuance: March 7, 2003.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment No.: 124.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2803).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2003.
No significant hazards consideration comments received: No.
GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear Station,
Unit 2, Dauphin County, Pennsylvania
Date of amendment request: November 14, 2002, supplemented by a
letter dated January 24, 2003, that supersedes previous applications
dated August 9, 2000, June 13, 2002.
Brief description of amendment request: The amendment revises TS
6.5.4 and 6.5.3 to eliminate the requirements for the Independent
Onsite Safety Review Group (IOSRG) which is not needed for safe
monitoring of TMI-2 based on consideration that the reactor has been
defueled to the extent reasonably achievable and the fuel shipped
offsite. The amendment also revises TS 6.4 to delete the requirements
for unit staff training that are outdated based on the adoption of a
systems approach to training consistent with 10 CFR 50.120, ``Training
and Qualification of Nuclear Power Plant Personnel.''
Date of issuance: March 5, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 59.
Facility Operating License No. DPR-73: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50955).
The November 14, 2002, application and supplemental letter dated
January 24, 2003, replace in their entirety the previous applications
dated August 9, 2000, June 13, 2002. The November 14, 2002, application
supplemented by the January 24, 2003, letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated March 5, 2003.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of application for amendment: July 23, 2002, as supplemented
November 15, 2002, and January 24, 2003.
Brief description of amendment: The amendment revises the Unit 2
reactor coolant system pressure-temperature curves in Technical
Specification (TS) Figures 3.4-2 and 3.4-3 and associated TS Bases. The
revised curves will bound operation of the unit for the remainder of
its current license duration and bound operation with planned license
amendments to increase the power level at which the unit is allowed to
operate.
Date of issuance: March 20, 2003.
Effective date: As of the date of issuance and shall be implemented
prior to startup from Unit 2 refueling outage 14.
Amendment No.: 255.
Facility Operating License No. DPR-74: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66010).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 20, 2003.
No significant hazards consideration comments received: No.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: January 13, 2000, and
supplemented by letters dated June 1, 2001, August 13, 2001, and
October 15, 2002.
Brief description of amendment: The amendment adds License
Condition 2.B.(9) to the MY license. This new license condition
incorporates the, Nuclear Regulatory Commission (NRC) approved,
``License Termination Plan Rev 3.'' (LTP), and associated addendum,
into the MY license and allows the licensee to make certain changes to
the approved LTP without prior NRC review and approval.
Date of issuance: February 28, 2003.
Effective date: Date of issuance; to be implemented within [30]
days from the date of issuance.
Amendment No.: 168.
[[Page 15770]]
Facility Operating License No. DPR-36: The amendment adds License
Condition 2.B.(9).
Date of initial notice in Federal Register: March 19, 2002.
The supplemental letters provided additional clarifying
information, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination published in the
Federal Register on March 19, 2002.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation Report dated February 28, 2003.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 26, 2002.
Brief description of amendment: The amendment revises Surveillance
Requirement (SR) 3.0.3 to extend the delay period, before entering a
Limiting Condition for Operation, following a missed surveillance. The
delay period is extended from the current limit of `` * * * up to 24
hours or up to the limit of the specified Frequency, whichever is
less'' to ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is greater.'' In addition, the following
requirement is added to SR 3.0.3: ``A risk evaluation shall be
performed for any Surveillance delayed greater than 24 hours and the
risk impact shall be managed.''
Date of issuance: March 6, 2003.
Effective date: March 6, 2003.
Amendment No.: 197.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75882).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 6, 2003.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: November 15, 2002, as supplemented by
letter dated February 24, 2003.
Brief description of amendment: The amendment revises the safety
limit minimum critical power ratio values in Technical Specification
2.1.1.2.
Date of issuance: March 17, 2003.
Effective date: March 17, 2003.
Amendment No.: 198.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78521).
The supplemental letter provided clarifying information that was
within the scope of the original Federal Register Notice (67 FR 78521)
and did not change the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2003.
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: July 26, 2002, as supplmented
December 19, 2002.
Brief description of amendment: The amendment revises Technical
Specification (TS) 1.0, ``Definitions,'' TS 2.1, ``Safety Limits,
Reactor Core,'' TS 2.3, ``Limiting Safety System Settings, Protective
Instrumentation,'' TS 3.1, ``Reactor Coolant System,'' TS 3.8,
``Refueling Operations,'' TS 3.10, ``Control Rod and Power Distribution
Limits,'' TS 6.9, ``Reporting Requirements,'' and their associated
Bases. These modifications allow the licensee to implement a Core
Operating Limits Report (COLR) by relocating cycle-specific, reactor
coolant system-related parameter limits from the TSs to the COLR. In
addition, the amendment makes administrative changes to the above TSs.
Date of issuance: March 11, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 165.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56322).
The supplemental information dated December 19, 2002, contained
clarifying information and did not change the scope of the July 26,
2002, application nor 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 March 11, 2003.
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: March 19, 2002, supplemented by
letters dated September 13 and October 21, 2002.
Brief description of amendment: The amendment revises the current
radiological consequence analyses for the Kewaunee Nuclear Power Plant
(KNPP) design-basis accidents to implement the alternate source term
(AST) as described in Regulatory Guide 1.183, ``Alternative
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors'' and Pursuant to 10 CFR 50.67, ``Accident
Source Term.''
Date of issuance: March 17, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 166.
Facility Operating License No. DPR-43: Amendment revised the
current radiological consequence analyses for the KNPP design-basis
accidents to implement the AST.
Date of initial notice in Federal Register: April 16, 2002 (67 FR
18646).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: January 4, 2002, as supplemented
January 9, 2003.
Brief description of amendment: The amendment adds a limiting
condition for operation of the mechanical vacuum pump instrumentation
to trip the pumps on indication of high radiation levels in the main
steam line and adds associated Surveillance Requirements.
Date of issuance: March 11, 2003.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No.: 143.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 19, 2002 (67
FR 7421).
The January 9, 2003, supplement contained clarifying information
and
[[Page 15771]]
did not change the staff's proposed finding of no significant hazards
consideration. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 11, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: August 20, 2002.
Brief description of amendment: The amendment modifies the diesel
generator action statements and surveillance requirements defined in
the plant's Technical Specifications, in order to reduce degradation of
the diesel generators associated with fast starting and rapid loading.
Date of issuance: March 17, 2003.
Effective date: March 17, 2003.
Amendment No.: 144.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61684).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2003.
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: March 20, 2002.
Brief description of amendment: The proposed amendment would change
Technical Specification (TS) Section 1.10, ``Definitions, Dose
Equivalent I-131,'' to allow the use of the thyroid dose conversion
factors listed in the International Commission on Radiological
Protection Publication No. 30 (ICRP-30), ``Limits for Intakes of
Radionuclides by Workers,'' 1979, in determining the iodine-131 dose
equivalent reactor coolant activity in TS Section 3/4.4.8 and in
calculating the radiological consequences from postulated design basis
accidents.
Date of issuance: March 6, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 162.
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53991).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 6, 2003.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: August 9 2002, as supplemented
by letters dated January 8 and February 6, 2003.
Brief description of amendments: The amendments revised the Updated
Final Safety Analysis Report to incorporate the Boiling Water Reactor
Vessel and Internals Project Integrated Surveillance for the
surveillance of the material capsules.
Date of issuance: March 10, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 237 and 179.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61684).
The supplements dated January 8 and February 6, 2003, provided
clarifying information that did not change the scope of the August 9,
2002, 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 March 10, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 24th day of March, 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-7489 Filed 3-31-03; 8:45 am]
BILLING CODE 7590-01-P