[Federal Register: September 2, 2003 (Volume 68, Number 169)]
[Notices]
[Page 52233-52240]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr02se03-140]
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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, August 8, 2003, through August 21, 2003.
The last biweekly notice was published on August 19, 2003, (68 FR
49812).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By October 2, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714, which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/.
If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the
[[Page 52234]]
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 hearingdocket@nrc.gov. 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 OGCMailCenter@nrc.gov. 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
pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: April 3, 2003.
Description of amendment request: The proposed amendment would
permit application of an alternative source term (AST) methodology,
according to Section 50.67, ``Accident source term,'' of title 10 of
the Code of Federal Regulations (10 CFR) with the exception that
Technical Information Document (TID) 14844, ``Calculation of Distance
Factors for Power and Test Reactor Sites,'' will continue to be used as
the radiation dose basis for equipment qualification. The proposed
amendment would include Technical Specifications (TS) and associated
Bases revisions to reflect implementation of AST assumptions; TS and
associated Bases revisions to increase main steam isolation valve
allowable leakage; TS and associated Bases revisions to decrease
allowed feedwater isolation valve leakage to allow margin to be used
for other release paths; TS and associated Bases revisions to delete
requirements for the main steam isolation valve leakage control system;
TS and associated Bases revisions to reflect requirements for
availability of Standby Liquid Control (SLC) System in Mode 3 and use
of the SLC System to buffer suppression pool pH to prevent iodine re-
evolution during a postulated radiological release; TS and associated
Bases revisions to reflect higher allowed charcoal adsorber
penetrations in laboratory testing; TS Bases revision to reflect an
increased allowed secondary containment drawdown time; TS Bases
revision to identify additional containment leakage exclusions from
La and exclusions from secondary containment bypass
allowances; additional allowance for filtered and unfiltered inleakage
into the control room envelope; and development of new offsite and
control room atmospheric dispersion factors calculated using site-
specific meteorology data collected between 2000 and 2002.
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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment implements alternative source term (AST)
assumptions in revisions to the analyses of the following limiting
design basis accidents at Clinton Power Station (CPS).
[sbull] Loss-of-Coolant Accident
[sbull] Main Steam Line Break Accident, and
[sbull] Control Rod Drop Accident
The AST does not require modification of the facility; rather,
once the occurrence of an accident has been postulated the new
source term is an input to evaluate the potential consequences. The
implementation of the AST has been evaluated in revisions to the
analyses of the limiting design basis accidents at CPS. Based upon
the results of these analyses, it has been demonstrated that, with
the requested changes, the dose consequences of these limiting
events is
[[Page 52235]]
within the regulatory guidance provided by the NRC for use with the
AST. This guidance is presented in 10 CFR 50.67 and associated
Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1.
The equipment affected by the revised operational conditions is
not considered an initiator to any previously analyzed accident and
therefore, inoperability of the equipment cannot increase the
probability of any previously evaluated accident. The radiological
consequences of the above design basis accidents have been evaluated
with applications of AST assumptions. The results conclude that the
radiological consequences remain within applicable regulatory
limits.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The application of AST does not affect the design, functional
performance or operation of the facility. Similarly, it does not
affect the design or operation of any structures, systems or
components involved in the mitigation of any accidents, nor does it
affect the design or operation of any component in the facility such
that new equipment failure modes are created.
As such the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Approval of the basis change from the original source term
developed in accordance with Technical Information Document (TID)
14844 to a new AST, as described in Regulatory Guide 1.183, is
requested. The results of the accident analyses revised in support
of the proposed changes, and the requested Technical Specification
changes, are subject to revised acceptance criteria. These analyses
have been performed using conservative methodologies as specified in
Regulatory Guide 1.183.
Safety margins and analytical conservatisms have been evaluated
and have been found acceptable. The analyzed events have been
carefully selected and margin has been retained to ensure that the
analyses adequately bound postulated event scenarios. The dose
consequences due to design basis accidents comply with the
requirements of 10 CFR 50.67 and the guidance of Regulatory Guide
1.183.
The margin of safety is considered to be that provided by
meeting the applicable regulatory limits. Relaxation of these
Technical Specification requirements results in an increase in dose
following certain design basis accidents. However, since the doses
following these design basis accidents remain within the regulatory
limits, there is not a significant reduction in a margin of safety.
The changes continue to ensure that the doses at the exclusion area
and low population zone boundaries, as well as the control room, are
within the corresponding regulatory limits.
Therefore, operation of CPS in accordance with the proposed
changes will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Deputy General Counsel
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station Unit No. 2, Oswego County, New York
Date of amendment request: August 15, 2003.
Description of amendment request: The licensee proposed to revise
the reactor coolant system pressure-temperature (P-T) limit curves
specified in Section 3.4.11, ``RCS [Reactor Coolant System] Pressure
and Temperature (P/T) Limits,'' of the Technical Specifications (TSs).
The proposed P-T limit curves will be based, in part, on an alternative
methodology and will be valid for 22 effective full-power years. The
alternative methodology, identified as American Society of Mechanical
Engineers Boiler and Pressure Vessel Code Case N-640, has been
previously approved for generic use by the Nuclear Regulatory
Commission (NRC).
The associated licensee-controlled TSs Bases pages would also be
changed to reflect the above TS changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the three standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The proposed changes, if approved by the NRC, will be made
in a manner such that conservatism is maintained through compliance
with applicable NRC regulations and guidance. No hardware design change
is involved with the proposed amendment, thus there will be no adverse
effect on the functional performance of any plant structure, system, or
component (SSC). All SSCs will continue to perform their design
functions with no decrease in their capabilities to mitigate the
consequences of postulated accidents. P-T limit curves were not
previously factored into the probability of accidents, nor were they
factored into scenarios of previously analyzed accidents. Accordingly,
the revised P-T limit curves will lead to no increase in the
consequences of an accident previously evaluated, and no increase of
the probability of an accident previously evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods the unit is operated. As a
result, all SSCs will continue to perform as previously analyzed by the
licensee, and previously evaluated and accepted by the NRC staff.
Therefore, the proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the licensee did not propose to
exceed or alter a design basis or safety limit, the proposed amendment
will not affect in any way the performance characteristics and intended
functions of any SSC. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: October 17, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification Table 3.3.1-2 by modifying a constant in
the variable
[[Page 52236]]
thermal margin/low pressure (TM/LP) trip equation. The proposed change
would reduce calculated values for the variable TM/LP trip equation.
The proposed equation constant value change results from improvements
in plant equipment used to establish the TM/LP trip setpoint.
Ultrasonic feedwater flow measurement devices, recently installed at
the Palisades Plant, result in less uncertainty applied in the
methodology used for determining core power level. Additionally, the
devices used to calculate the TM/LP trip setpoint have previously been
replaced with devices having less uncertainty. These reduced
uncertainties, when combined using the NRC-endorsed methodology
described in ANSI/ISA-S67.04-1994, ``Setpoints for Nuclear Safety-
Related Instrumentation,'' result in a reduction in the constant (bias
term) used to calculate the TM/LP trip setpoint.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following evaluation supports the finding that operation of
the facility in accordance with the proposed change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment does not involve operation of any
required structures, systems or components (SSCs) in a manner or
configuration different from those previously recognized or
evaluated. The methodology that was used in determining the
recommended change in the constant follows Nuclear Regulatory
Commission endorsed standard ANSI/ISA-S67.04-1994, ``Setpoints for
Nuclear Safety-Related Instrumentation.'' The probability of an
accident previously evaluated will not be increased since the
proposed change to the constant value in the Thermal Margin/Low
Pressure (TM/LP) trip equation maintains all necessary
considerations in the development of uncertainties.
The consequences of an accident previously evaluated will not be
increased since the reactor is still protected from violating the
TM/LP trip setpoint used in the safety analysis for the Palisades
Nuclear Plant.
Therefore, operation of the facility in accordance with the
proposed change to the Technical Specifications would 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 accident previously evaluated.
The proposed change to the constant value for the TM/LP trip
equation in the Technical Specifications would not change or add a
system function. The proposed amendment does not involve operation
of any required SSCs in a manner or configuration different from
those previously recognized or evaluated. No new failure mechanisms
will be introduced by the change being requested.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to the constant value for the TM/LP trip
equation in the Technical Specifications accounts for all
uncertainties that affect the TM/LP trip setpoint. The revised TM/LP
trip setpoint will continue to assure that the acceptance criteria
established in the safety analysis will be met.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: L. Raghavan.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: July 24, 2003.
Description of amendment requests: The proposed change will revise
Technical Specification (TS) Section 3.8.4, ``DC Sources--Operating'';
TS Section 3.8.5, ``DC Sources--Shutdown''; and TS Section 3.8.6,
``Battery Cell Parameters.'' The proposed change will also add a new
section to TS 5.5, ``Programs and Manuals'' for the maintenance and
monitoring of the station safety-related batteries that is based on the
recommendations of the Institute of Electrical and Electronics
Engineers (IEEE) Standard 450-1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change affects Technical Specification (TS)
sections 3.8.4 ``DC Sources--Operating,'' TS 3.8.5 ``DC Sources--
Shutdown,'' TS 3.8.6 ``Battery Cell Parameters,'' and TS
Administrative Controls section 5.5.
The proposed change restructures the TS for the direct current
(DC) electrical power subsystem and adds new Conditions and Required
Actions with increased Completion Times to address battery charger
inoperability. Neither the DC electrical power subsystem nor
associated battery chargers are initiators of any accident sequence
analyzed in the Final Safety Analysis Report Update (FSARU).
Operation in accordance with the proposed TS ensures that the DC
electrical power subsystem is capable of performing its function as
described in the FSARU. Therefore the mitigating functions supported
by the DC electrical power subsystem will continue to provide the
protection assumed by the analysis.
The relocation of preventive maintenance surveillances, and
certain operating limits and actions to a newly-created, licensee-
controlled TS 5.5.17, ``Battery Monitoring and Maintenance
Program,'' will not challenge the ability of the DC electrical power
subsystem to perform its design function. The maintenance and
monitoring required by current TS, which are based on industry
standards, will continue to be performed. In addition, the DC
electrical power subsystem is within the scope of 10 CFR 50.65,
``Requirements for monitoring the effectiveness of maintenance at
nuclear power plants,'' which will ensure the control of maintenance
activities associated with the DC electrical power subsystem.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve any physical alteration of
the units. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints at which protective or mitigating actions are
initiated that are affected by the proposed changes. The operability
of the DC electrical power subsystems in accordance with the
proposed TS is consistent with the initial assumptions of the
accident analyses and is based upon meeting the design basis of the
plant. The proposed change will not alter the manner in which
equipment operation is initiated, nor will the functional demands on
credited equipment be changed. No alteration in the operating
procedures, which ensure the unit remains within analyzed limits, is
proposed, and no change is being made to procedures relied upon to
respond to an off-normal event. As such, no new failure modes are
being introduced. The proposed change does not alter assumptions
made in the safety analyses.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
[[Page 52237]]
The proposed change will not adversely affect operation of plant
equipment and will not result in a change to the setpoints at which
protective actions are initiated. Sufficient DC capacity to support
operation of mitigation equipment is ensured. The changes associated
with the new battery maintenance and monitoring program will ensure
that the station batteries are maintained in a highly reliable
manner. The equipment fed by the DC electrical system will continue
to provide adequate power to safety-related loads in accordance with
analysis assumptions.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
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 (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 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 email to pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: December 20, 2002, as
supplemented by letter dated May 30, 2003.
Brief description of amendment: The amendment approves changes to
the Clinton facility as described in the Updated Safety Analysis
Report. The amendment modifies the basis for compliance with the
requirements of Appendix H to title 10 of the Code of Federal
Regulations part 50 (appendix H to 10 CFR part 50), ``Reactor Vessel
Material Surveillance Program Requirements,'' by approving
implementation of the Boiling-Water Reactor Vessel and Internals
Project reactor pressure vessel integrated surveillance program.
Date of issuance: August 12, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 157.
Facility Operating License No. NPF-62: The amendment approved
revisions to the Updated Safety Analysis Report.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5669). The supplemental letter 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 August 12, 2003.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: September 30, 2002, as
supplemented by letter dated March 19, 2003.
Brief description of amendment: The amendment revised Technical
Specification Section 6.8.5, ``Reactor Building Leakage Rate Testing
Program,'' to reflect a one-time deferral of the scheduled performance
of the next Type A Containment Integrated Leak Rate Test from October,
2003, to no later than September 2008.
Date of issuance: August 14, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 244.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68730). The supplement provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 14, 2003.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3 Maricopa County, Arizona
Date of application for amendments: April 25, 2003.
Brief description of amendments: The amendments revise Section 5.3,
``Unit Staff Qualifications,'' of the Technical Specifications to state
new education and experience eligibility requirements for operator
license applicants. As stated in the letter dated April 25, 2003, the
new requirements are outlined by the National Academy for Nuclear
Training in its ``Guidelines for Initial Training and Qualification of
Licensed Operators,'' which were issued January 2000.
Date of issuance: August 13, 2003.
Effective date: August 13, 2003, and shall be implemented within 90
days of the date of issuance.
Amendment Nos.: Unit 1-148, Unit 2-148, Unit 3-148.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
[[Page 52238]]
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34662). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 13, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-003, Indian Point
Nuclear Generating Station, Unit 1
Date of amendment request: May 30, 2002.
Brief description of amendment: It would revise the Indian Point
Nuclear Generating Station, Unit 1 (IP1) Technical Specifications (TSs)
to facilitate the Indian Point Generating Station, Unit 2 (IP2)
transition to the Improved TSs. The amendment also revises the
requirements of the ``Order Approving Decommissioning Plan and
Authorizing Decommissioning of Facility'' \1\ to ensure compliance with
the current requirements of 10 CFR 50.59 and 10 CFR 50.83. It also
revises the expiration date of Provisional Operating License No. DPR-5
for IP1 to be current with the expiration date for the Facility
Operating License No. DPR-26 for IP2.
---------------------------------------------------------------------------
\1\ NRC letter to Consolidated Edison, ``Order to Authorize
Decommissioning and Amendment No. 45 to License No. DPR-5 for Indian
Point Unit 1 (TAC No. M59664),'' dated January 31, 1996.
---------------------------------------------------------------------------
Date of issuance: August 11, 2003.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No: 52.
Provisional Operating License No. DPR-5: The amendment revised the
Technical Specifications, and made changes to and revised the
expiration date for IP1 Provisional Operating License DPR-5.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45564).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 11, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: September 19, 2002, as
supplemented by letters dated January 8, May 22, and July 1, 2003.
Brief description of amendment: The amendment extends the allowable
outage time for the emergency diesel generators from 72 hours to a
maximum of 14 days.
Date of issuance: August 8, 2003.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 249.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68733). The January 8, May 22, and July 1, 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 August 8, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: August 7, 2002, as supplemented
by your letters dated February 28, and May 27, 2003.
Brief description of amendments: The amendments revised the
limiting condition for operation, the associated Conditions and
Required Actions of Technical Specification (TS) 3.7.1, ``Main Steam
Safety Valves (MSSVs),'' and the values in Table 3.7.1-1, ``Operable
Main Steam Safety Valves versus Applicable Power in Percent of Rated
Thermal Power,'' by requiring five MSSVs per steam generator to be
operable consistent with the accident analyses assumptions. The
amendments also modify the associated Required Actions of TS 3.7.1 by
adding a requirement to reduce the Power Range Neutron Flux-High
reactor trip setpoint when one or more steam generators with one or
more MSSVs are inoperable.
Date of issuance: August 12, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 133/133, 128/128.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61681). 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 amendments is contained
in a Safety Evaluation dated August 12, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: December 20, 2002, as
supplemented by letters dated May 30, and June 27, 2003.
Brief description of amendments: The amendments approve changes to
the LaSalle County Station facility as described in the Updated Final
Safety Analysis Report. The amendments modify the basis for compliance
with the requirements of appendix H to title 10 of the Code of Federal
Regulations part 50 (appendix H to 10 CFR part 50), ``Reactor Vessel
Material Surveillance Program Requirements,'' by approving
implementation of the Boiling-Water Reactor Vessel and Internals
Project reactor pressure vessel integrated surveillance program.
Date of issuance: August 13, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 160/146.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
approve revisions to the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5669). 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 amendments is contained
in a Safety Evaluation dated August 13, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: November 30, 2001.
Brief description of amendment: This amendment revises Technical
Specification 3/4.4.4, ``Reactor Coolant System--Pressurizer,'' to
adopt a new pressurizer high-level limit and to revise the required
action when the pressurizer is inoperable.
Date of issuance: August 12, 2003.
[[Page 52239]]
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 255.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37578). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 12, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: May 21, 2003.
Brief description of amendment: This amendment relocates to the
Technical Requirements Manual the Technical Specification surveillance
requirement pertaining to flow balance testing of the emergency core
cooling system (ECCS) high pressure injection and low pressure
injection subsystems following system modifications that alter
subsystem flow characteristics. Also, the amendment adds an ECCS pump
operability requirement to the Technical Specifications.
Date of issuance: August 12, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 256.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34669). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 12, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: May 30, 2003.
Brief description of amendment: The amendment deletes technical
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminates the requirements to have and maintain the post accident
sampling system (PASS) at the Duane Arnold Energy. The amendment also
addresses related changes to TS 5.5.2, ``Primary Coolant Sources
Outside Containment.''
Date of issuance: August 8, 2003.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment No.: 252.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40713).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 8, 2003.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 8, 2002.
Brief description of amendment: The October 8, 2002, submittal
proposed the following: (1) The use of a pressure temperature limits
report (PTLR), (2) change the minimum boltup temperature, (3) revise
the low temperature overpressure protection (LTOP) methodology and
analysis, (4) perform the LTOP analyses ``in-house,'' (5) change the
LTOP enable temperature, (6) modify TS 2.10.1 to exactly specify the
reactor coolant system (RCS) temperature at which the reactor can be
made critical, and (7) add a TS for a maximum pressure value for the
safety injection tanks. This amendment approves the use of a PTLR for
the Fort Calhoun Station. As such TS Figure 2-1 (RCS Pressure--
Temperature Limits for Heatup, Cooldown, and In-service Test) will be
relocated into Figure 5-1 of the PTLR. In addition, the following TSs
were either modified or added for the implementation of the PTLR:
define the PTLR in Definitions; TS 2.1.1(8); TS 2.1.1(11); TS 2.1.2 and
2.1.2 References; TS 2.1.6(4); TS 2.3(1)(c); TS 2.3(3); TS 2.3
References; TS 2.10.1; Table 3-5, item 23, TS 3.3(1)(c); and TS 5.9.6.
The following TS Bases sections were modified to reflect the
implementation of the PTLR: TS 2.1.1, TS 2.1.2, and TS 2.10.1.
Date of issuance: August 15, 2003.
Effective date: August 15, 2003. The amendment shall be implemented
within 30 days from the date of issuance, including submitting the
first Pressure Temperature Limits Report to the NRC Document Control
Desk with copies to the Region IV Regional Administration and Resident
Inspector.
Amendment No.: 221.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37579). The April 10, June 4, July 31, and August 5, 2003, supplemental
letters provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 15, 2003.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 8, 2002, as supplemented by
letters dated April 11 and May 21, 2003.
Brief description of amendment: The amendment grants a one-time
five-year extension to the current ten-year test interval for the
containment integrated leak rate testing.
Date of issuance: August 15, 2003.
Effective date: August 15, 2003, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 220.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68742). The April 11 and May 21, 2003, supplemental letters provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 15, 2003.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: June 5, 2003.
Brief description of amendments: The amendments extend from 1 hour
to 24 hours the completion time for Condition B of Technical
Specification 3.5.1, which defines requirements for the restoration of
an emergency core cooling system accumulator when it has been declared
inoperable for a reason other than boron concentration.
Date of issuance: August 15, 2003.
Effective date: August 15, 2003, and shall be implemented within 60
days from the date of issuance.
Amendment Nos.: Unit 1--160; Unit 2--161.
[[Page 52240]]
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40716). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 15, 2003.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 31, 2003.
Brief description of amendments: The amendments replace ``Central
Power and Light Company (CPL)'' with ``AEP Texas Central Company''
throughout the Operating License of each unit.
Date of issuance: August 11, 2003.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment Nos.: Unit 1--155; Unit 2--143.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34673). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 11, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: December 13, 2002, as
supplemented May 19 and July 11, 2003.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.7.2.12, ``Steam Generator (SG) Tube Surveillance
Program.'' The revised TS allows the use of Westinghouse leak-limiting
Alloy 800 sleeves to repair defective SG tubes as an alternative to
plugging the tube.
Date of issuance: August 15, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 44.
Facility Operating License No. NPF-90: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003
(68FR12958). The supplemental letters provided clarifying information
that did not expand the scope of the original request and did not
change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 15, 2003.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: November 5, 2002.
Brief Description of amendments: These amendments delete the
requirement to perform a 15-minute degassed beta and gamma activity
test of the secondary coolant and require that the dose equivalent I-
131 analysis be performed on a more conservative monthly basis.
Date of issuance: August 15, 2003.
Effective date: August 15, 2003.
Amendment Nos.: 234 and 233.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78525). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 15, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 25th day of August, 2003.
For the Nuclear Regulatory Commission.
Eric J. Leeds,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 03-22106 Filed 8-29-03; 8:45 am]
BILLING CODE 7590-01-P