[Federal Register Volume 68, Number 160 (Tuesday, August 19, 2003)]
[Notices]
[Pages 49812-49825]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-20839]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Background
Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Public Law 97-415 revised section 189 of the Atomic Energy Act
of 1954, as amended (the Act), to require the Commission to publish
notice of any amendments issued, or proposed to be issued, under a new
provision of section 189 of the Act. This provision grants the
Commission the authority to issue and make immediately effective any
amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 25, 2003, through August 7, 2003. The
last biweekly notice was published on August 5, 2003 (68 FR 46239).
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.
[[Page 49813]]
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 September 18, 2003, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2.
Interested persons should consult a current copy of 10 CFR 2.714, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions,
[[Page 49814]]
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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: July 14, 2003.
Description of amendments request: The proposed amendment would
revise the Technical Specifications to eliminate Surveillance
Requirement (SR) 3.6.6.8. This SR is a 10-year flow test to verify that
the containment spray nozzles are unobstructed.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change eliminates the surveillance requirement to
verify that the Containment Spray System spray nozzles are
unobstructed every ten years. The spray nozzles are not initiators
of any previously analyzed accident. Therefore, this proposed change
does not increase the probability of any accident previously
evaluated.
The spray nozzles are assumed in the accident analysis to
mitigate design basis accidents. Calvert Cliffs' system design
Foreign Material Exclusion practices during maintenance and material
accountability following maintenance, and post-maintenance testing
practices ensure that the system is free of foreign material that
could significantly reduce its ability to perform its intended
function. These controls are considered adequate to ensure continued
operability of the spray system. Since the system will be able to
perform its accident mitigation function, the consequences of
accidents previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed change eliminates the surveillance requirement to
verify that the Containment Spray System spray nozzles are
unobstructed every ten years. The proposed change does not introduce
a new method of plant operation, does not involve a physical
modification to the plant, nor does it introduce any accident
initiators.
Therefore, the proposed change does not create the possibility
of a new or different [kind] of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
The margin of safety in this case is the assurance of
operability of the Containment Spray System. Calvert Cliffs' system
design, Foreign Material Exclusion practices during maintenance and
material accountability following maintenance, and post-maintenance
testing practices ensure that the system is free of foreign material
that could significantly reduce its ability to perform its intended
function. These requirements, along with the remote physical
location and the simple construction of the spray nozzles, provide
assurance that the nozzles will remain operable.
Therefore, this 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
amendments request involves no significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, Counsel,
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor,
Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: May 29, 2003.
Description of amendments request: The proposed license amendments
request approval to remove the current Brunswick Steam Electric Plant
(BSEP) reactor material specimen surveillance schedule from the Updated
Final Safety Analysis Report and specify that BSEP, Units 1 and 2, will
participate in an integrated surveillance program (ISP) developed by
the Boiling Water Reactor Vessel and Internals Project.
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 adopts an integrated surveillance program
(ISP) for reactor vessel material specimen surveillances. The ISP
ensures that the reactor pressure vessel will continue to meet all
applicable fracture toughness requirements. No physical changes to
the facilities will result from the proposed change. The initial
conditions and methodologies used in accident analyses remain
unchanged. The proposed change does not revise the design
assumptions for systems or components used to mitigate the
consequences of accidents. The accident analyses results are not
affected by this proposed change. 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 adopts an integrated surveillance program
(ISP) for reactor vessel material specimen surveillances. The ISP
ensures that the reactor pressure vessel will continue to meet all
applicable fracture toughness requirements. No physical changes to
the facilities will result from the proposed change. The proposed
change does not affect the design or operation of any system,
structure, or component in the facilities. The safety functions of
the related systems, structures, or components are not changed in
any manner, nor is the reliability of any system, structure, or
component reduced. The change does not affect the manner by which
the facilities are operated and does not change any facility,
structure, or component. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change has no impact on the margin of safety of any
Technical Specification. There is no impact on safety limits or
limiting safety system settings. The proposed change does not affect
any plant safety parameters or setpoints. No physical or operational
changes to the facilities will result from the proposed change.
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
[[Page 49815]]
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.
NRC Section Chief: Allen G. Howe.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: June 24, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV)
Vent and Drain Valves,'' to allow a vent or drain line with one
inoperable valve to be isolated instead of requiring the valve to be
restored to Operable status within 7 days.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments to revise the action for one or more SDV vent or drain lines
with an inoperable valve, including a model safety evaluation and model
no significant hazards consideration (NSHC) determination, using the
consolidated line-item improvement process (CLIIP). The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on April 15, 2003 (68 FR 18295). The licensee affirmed the
applicability of the NSHC determination below in its application dated
June 24, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with one valve inoperable instead or requiring the valve to be
restored to operable status within 7 days. With one SDV vent or
drain valve inoperable in one or more lines, the isolation function
would be maintained since the redundant valve in the affected line
would perform its safety function of isolating the SDV. Following
the completion of the required action, the isolation function is
fulfilled since the associated line is isolated. The ability to vent
and drain the SDVs is maintained and controlled through
administrative controls. This requirement assures the reactor
protection system is not adversely affected by the inoperable
valves. With the safety functions of the valves being maintained,
the probability or consequences of an accident previously evaluated
are not significantly increased.
Criterion 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 a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDVs is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of an SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: November 14, 2002.
Description of amendment request: Pursuant to Title 10 of the Code
of Federal Regulations (10 CFR) Sec. 50.90, Duke Energy Corporation
requested an amendment to the McGuire Nuclear Station Facility
Operating Licenses and Technical Specifications (TS). The proposed
change would revise TS 3.3.2, Engineered Safety Features Actuation
System Instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following discussion is a summary of the evaluation of the
changes contained in this proposed license amendment against the
three required standards of 10 CFR 50.92(c). A no significant
hazards consideration is indicated if 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.
First Standard. Implementation of this amendment would not
involve a significant increase in the probability or consequences of
an accident previously evaluated. Implementation of the changes
contained in this amendment will have no effect on accident
probabilities or consequences. The proposed changes apply to
Technical Specifications 3.3.2, Engineered Safety Features Actuation
System, and the equipment referenced in this Technical Specification
are not accident initiating equipment. Therefore, there will be no
impact on any accident probabilities caused by the NRC approval of
this license amendment request. Additionally, since the design of
the equipment is not being adversely modified by these proposed
changes, there will be no impact on any accident consequences.
Second Standard. Implementation of this amendment would not
create the possibility of a new or different kind of accident from
any accident previously evaluated. No new accident causal mechanisms
will be created as a result of the NRC approval of this license
amendment request. No changes are being made to the plant which will
introduce any new accident causal mechanism. This amendment does not
impact any plant systems that are accident initiators; therefore, no
new accident types are being created.
Third Standard. Implementation of this amendment would not
involve a significant reduction in a margin of safety. Margin of
safety is related to the confidence in the ability of the fission
product barriers to perform their design functions during and
following an accident situation. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
The performance of these fission product barriers will not be
impacted by implementation of this amendment. The equipment
referenced in the proposed change to Technical Specification 3.3.2
will remain capable of performing as designed. No safety margins
will be impacted.
Conclusion. Based upon the preceding discussion, Duke Energy
Corporation has concluded that this proposed license amendment 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.
[[Page 49816]]
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
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: July 10, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to eliminate requirements that are
no longer applicable due to the completion of the automatic feedwater
system modifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.91, Duke Energy Corporation (Duke) has
made the determination that this amendment request involves a No
Significant Hazards Consideration by applying the standards
established by the NRC regulations 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: The proposed
change to the Oconee Technical Specifications removes obsolete
requirements associated with the Main Steam Line Break (MSLB)
detection circuitry that are no longer necessary because of the
completion of the Automatic Feedwater Isolation System (AFIS)
modification on all three Oconee Units. AFIS replaced the MLSB
detection system. As such, the proposed change is administrative. No
actual plant equipment, operating practices, or accident analyses
are affected by this change. Therefore, the 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: The
proposed change to the Oconee Technical Specifications removes
obsolete requirements associated with the MSLB detection circuitry
that are no longer necessary because of the completion of the AFIS
modification on all three Oconee Units. AFIS replaced the MLSB
detection system. As such, the proposed change is administrative. No
actual plant equipment, operating practices, or accident analyses
are affected by this change. No new accident causal mechanisms are
created as a result of this change. The proposed change does not
impact any plant systems that are accident initiators; neither does
it adversely impact any accident mitigating systems. 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 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. The proposed change eliminates obsolete
requirements and is administrative in nature. 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. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois and
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of application for amendment request: October 10, 2002, as
supplemented March 21 and March 28, 2003.
Description of amendment request: The proposed amendment revises
the licensing bases and Technical Specifications by utilizing an
alternative source term in the design-basis radiological analyses in
accordance with 10 CFR 50.67, with the exception that Technical
Information Document 14844 will continue to be used as the radiation
dose basis for equipment qualification.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The implementation of alternative source term (AST) assumptions
has been evaluated in revisions to the analyses of the following
limiting design basis accidents at Dresden Nuclear Power Station
(DNPS) and Quad Cities Nuclear Power Station (QCNPS):
Loss-of-Coolant Accident,
Main Steam Line Break Accident,
Fuel Handling Accident, and
Control Rod Drop Accident.
Based upon the results of these analyses, it has been
demonstrated that, with the requested changes, the dose consequences
of these limiting events is 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.
Requirements for secondary containment operability, secondary
containment isolation valves, the Standby Gas Treatment (SGT)
System, the Control Room Emergency Ventilation (CREV) System, and
the Control Room Emergency Ventilation Air Conditioning (AC) System
during movement of irradiated fuel assemblies that have decayed at
least 24-hours and during core alterations are being eliminated.
This is acceptable because, with the application of AST, none of
these systems are credited in mitigating the consequences of a fuel
handling accident after a 24-hour decay period.
The proposed change also increases the maximum allowable primary
containment leakage and the maximum allowable main steam isolation
valve leakage limits. This is acceptable due to the new assumptions,
used in calculating control room and offsite dose following a design
basis loss-of-coolant accident, related to application of AST.
The proposed changes do not affect the design or operation of
the facility; rather, once the occurrence of an accident has been
postulated, the new source term is an input to evaluate the
consequence. The radiological consequences of the above design basis
accidents have been evaluated with application 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. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
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. The proposed change does not involve a significant reduction
in a margin of safety.
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.
[[Page 49817]]
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 DNPS and QCNPS 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: 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. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 27, 2003.
Description of amendment request: The proposed amendments revise
Technical Specification 4.0.5.f and associated Bases, and Bases Section
3/4.4.8, with regard to the commitment to perform piping inspections in
accordance with Generic Letter 88-01, by adding the words ``or in
accordance with alternate measures approved by the NRC staff.''
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 U.S. Nuclear Regulatory Commission (NRC) staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below.
1. Do the proposed amendments involve a significant increase in the
probability or consequences of an accident previously evaluated?
No physical changes to the facilities will result from the proposed
changes. The initial conditions and methodologies used in accident
analyses remain unchanged. The proposed changes do not revise or alter
the design assumptions for systems or components used to mitigate the
consequences of accidents. Thus, accident analyses results are not
affected by these changes. Therefore, the proposed amendments do not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed amendments create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes do not affect the design or operation of any
system, structure, or component in the plants. No new or different type
of equipment will be installed by these proposed changes. Therefore,
the proposed amendments do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Do the proposed amendments involve a significant reduction in a
margin of safety?
The changes do not affect any plant safety parameters or setpoints.
No physical or operational changes to the facility will result from the
proposed changes. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
Based on the NRC staff's 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: Mr. Edward Cullen, Vice President & General
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett
Square, PA 19348.
NRC Section Chief: James W. Clifford.
Nuclear Management Company (NMC), LLC, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: February 11, 2003, as supplemented July
16, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 5.5.9, ``Ventilation Filter Testing
Program (VFTP),'' by (1) incorporating filter test face velocity limits
for the control room special ventilation system, auxiliary building
special ventilation system, spent fuel pool special and inservice purge
ventilation system, and shield building ventilation system; and (2)
making editorial changes. The proposed amendments would also delete the
additional conditions in Appendix B of the Operating Licenses which
require the licensee to complete an evaluation of the maximum test face
velocity for the ventilation systems in TS 5.5.9. The additional
conditions would also require the licensee to submit a license
amendment request for a TS amendment to specify the maximum test face
velocity if the maximum actual face velocity is greater than 110
percent of 40 feet per minute.
In its July 16, 2003, supplemental letter, NMC withdrew the portion
of its original request to revise the penetration and system bypass
limit from 0.05 percent to 0.5 percent for the ventilation systems. The
proposed amendments were previously noticed in the Federal Register on
April 15, 2003 (68 FR 18279).
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:
Addition of Filter Test Face velocities
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes to add filter test face
velocity minimum values for the control room special ventilation
system, auxiliary building special ventilation system, spent fuel
pool special and inservice purge ventilation system and shield
building ventilation system. These ventilation systems are included
in the plant design to mitigate accident consequences and are not
assumed accident initiators, thus, this change does not involve a
significant increase in the probability of an accident. This change
will assure that the subject ventilation systems will perform within
their intended design ranges thus, this change assures that the
consequences of an accident are not increased.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. The malfunction of safety related equipment,
assumed to be operable in the accident analyses, would not be caused
as a result of the proposed Technical Specification change. No new
failure mode has been created and no new equipment performance
burdens are imposed. Therefore the possibility of a new
[[Page 49818]]
or different kind of accident from those previously analyzed has not
been created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes to add filter test face
velocity minimum values for the control room special ventilation
system, auxiliary building special ventilation system, spent fuel
pool special and inservice purge ventilation system and shield
building ventilation system. These additional Technical
Specification limits on system performance assures these ventilation
systems are tested and maintained within their designed function
limits and may increase the margin of safety for these systems.
Therefore this change does not involve a significant reduction in
the margin of safety.
Editorial and administrative changes
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes editorial changes to
Technical Specification Section 5.5.9, including replacement of
ventilation system names with abbreviations and miscellaneous
changes associated with addition of a new paragraph to this section,
and proposes an administrative change to delete the Operating
License Additional Condition for each unit that relates to NRC
Generic Letter 99-02. Since these changes are editorial or
administrative, they do not change any plant operating limits or
technical requirements. Therefore these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. The malfunction of safety related equipment,
assumed to be operable in the accident analyses, would not be caused
as a result of the proposed technical specification change. No new
failure mode has been created and no new equipment performance
burdens are imposed. Therefore, the possibility of a new or
different kind of accident from those previously analyzed has not
been created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes editorial changes to
Technical Specification Section 5.5.9, including replacement of
ventilation system names with abbreviations and miscellaneous
changes associated with addition of a new paragraph to this section,
and proposes an administrative change to delete the Operating
License Additional Condition for each unit that relates to NRC
Generic Letter 99-02. Since these changes are editorial or
administrative, they do not change any plant operating limits or
technical requirements. Therefore these changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 25, 2003.
Description of amendment request: The proposed amendment modifies
Technical Specification 2.1.4, ``Reactor Coolant System (RCS) Leakage
Limits.'' The proposed amendment will: (1) Add a requirement for no RCS
pressure boundary leakage, (2) combine the existing RCS leakage limits
into a format similar to the Improved Standard Technical Specification
(ISTS), and (3) replace the existing basis associated with this
specification with a basis similar in format and content of the ISTS.
The proposed changes will assure that the design criteria of no RCS
pressure boundary leakage is maintained and bring the Fort Calhoun
Station, Unit 1 (FCS) RCS leakage specifications into alignment with
the Improved Standard Technical Specifications. This amendment is
modeled after the Improved Standard Technical Specifications.
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 to Technical Specifications 2.1.4 establish
a limit on reactor coolant system pressure boundary leakage and
provide an allowed outage time and actions required for restoring
operability. The proposed Technical Specifications address the
regulatory requirements for equipment required for FCS Design
Criterion 16 (similar to 10 CFR 50 GDC [General Design Criterion]
30). The change will ensure that proper Limiting Conditions for
Operation are entered for equipment or functional inoperability.
There are no physical alterations being made to the reactor coolant
system or related systems. Therefore, the proposed changes do 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 changes will not result in any physical alterations
to the reactor coolant system, any plant configuration, systems,
equipment, or operational characteristics. There will be no changes
in operating modes, or safety limits, or instrument limits. With the
proposed changes in place, Technical Specifications will retain
requirements for the reactor coolant system. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes clarify the regulatory requirements for the
reactor coolant system as defined by FCS Design Criterion 16
(similar to 10 CFR 50 GDC 30). The times established are within
those invoked by the present Technical Specifications or equal to
those previously reviewed and approved for use by the NRC. The
proposed changes will not alter any physical or operational
characteristics of the reactor coolant system and associated systems
and equipment. Therefore, the proposed changes do 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: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW, Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 25, 2003.
Description of amendment request: The proposed amendment modifies
Technical Specifications (TS) 3.0.2, Table 3-2, Table 3-5, 3.6, 3.7,
3.8, and the Definitions Section. This proposed change provides a risk-
informed alternative to the existing surveillance interval for the
integrated engineered safety features (ESF) and loss-of-offsite power
(LOOP) testing required to be performed on each ESF equipment train
each outage. The proposed change modifies the surveillance interval
requirement for these refueling interval surveillance requirements to
go to a staggered test-basis scheme. Using a staggered test basis, only
one train would be tested each refueling outage. This amendment is
modeled after the
[[Page 49819]]
Improved Standard Technical Specifications (ISTS) and is based on a
study conducted by the Westinghouse Electric Company, on behalf of the
Combustion Engineering Owners Group (CEOG) in Topical Report WCAP-
15830-P, ``Staggered Integrated ESF Testing,'' and Technical
Specification Task Force (TSTF) 450.
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 only the Frequency at which
integrated ESF testing should be performed. This testing provides
assurance that the integrated ESF response will occur as assumed in
the accident analyses. Testing of the components will continue to be
performed as currently specified in the Technical Specifications.
The only change will be for the integrated test. This test will
continue to be performed on each train of ESF equipment, however, it
will be performed on a Staggered Test Basis. This means that the
testing will be less frequent than currently required. However,
testing seldom shows failure of the equipment to perform its safety
function. Because of the complexity of performing the test, the test
is most likely to be repeated for some discrepancy in the set up of
the test. The detailed risk review and assessment of a longer test
interval shows that the change in risk is low or unchanged for
equipment covered by the topical report. Licensees will provide
acceptable risk reviews for plant specific equipment.
This test does not increase the probability of an accident
previously evaluated because it is not a precursor to an accident.
In addition, the test is performed in a shutdown mode, where these
types of accidents are not assumed to occur. The proposed change
also does not increase the consequences of an accident previously
evaluated because the equipment is still demonstrated to perform its
safety function in an integrated manner. One complete train of
equipment will be tested every refueling interval for each train.
Successful completion of the test is still required.
Therefore, the proposed change will 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 affects only the Frequency at which
integrated ESF testing should be performed. All more frequently
performed testing is unaffected by this proposed change. No changes
are being made to the equipment or to the method of equipment
operation as a result of this change. No changes are being made to
the tests addressed by this proposed change except the frequency.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change affects only the surveillance interval at
which integrated ESF testing should be performed. It does not impact
safety system design criteria; safety system setpoint calculations
or assumptions made in the safety analyses. All of the affected
systems will continue to perform their safety functions as designed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW, Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: July 28, 2003.
Description of amendment requests: The proposed amendments would
revise Technical Specification 3.3.1, ``RPS Instrumentation--
Operating,'' and 3.3.5, ``ESFAS Instrumentation.'' Specifically, the
proposed changes would replace the requirement for the Steam Generator
Pressure--Low allowable value from its current value of 729 psia to a
revised value of 717 psia.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Accidents evaluated in Chapter 15 of the Updated Final Safety
Analysis Report use an analytical value of 675 psia for Steam
Generator Pressure--Low and for the Main Steam Isolation Signal/
Emergency Feedwater Actuation Signal, which is the basis for the
proposed change to the allowable value. The current and proposed
Allowable Values are 729 psia and 717 psia respectively, which means
that a 12 psi reduction in margin between the Allowable Value and
the Analytical Value is being proposed. Since the Trip Setpoint may
not be below 717 psia (it would be at 725 psia as required by the
supporting calculation), the proposed reduction in margin between
the Allowable Value and the Analytical Value does not involve a
significant increase in the consequences of an accident previously
evaluated.
The proposed amendment has no effect on the probability of
occurrence of accidents evaluated in Chapter 15 of the Updated Final
Safety Analysis Report.
Therefore, the proposed change does 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.
There will be no change to the design basis of the plant. There
are no new anticipated operational occurrences, or design basis
accidents. No changes to any other analytical limits are being made.
The current Analytical Value for Steam Generator Pressure--Low is
being retained, and no changes to any of the assumptions in the
accident analyses are being proposed.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The change in allowable value will not adversely affect the
design analysis. The plant would trip on Steam Generator Pressure--
Low at values at, or above, the analysis limit. The proposed change
in the Allowable Value does not involve any change to the Analytical
Value, so that the design bases limit is maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, Southern California Edison concludes that
the proposed amendments present no significant hazards consideration
under the standards set forth in 10 CFR 50.92(c), and, accordingly,
a finding of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
[[Page 49820]]
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: July 21, 2003.
Brief description of amendments: The proposed change would revise
Technical Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance
Program'', to allow the use of Westinghouse Electric LLC (Westinghouse)
leak limiting Alloy 800 sleeves to repair defective SG tubes as an
alternative to plugging these tube.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), Section 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 Westinghouse Alloy 800 leak limiting repair sleeves are
designed using the applicable American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code [Code] and,
therefore, meet the design objectives of the original steam
generator tubing. The applied stresses and fatigue usage for the
repair sleeves are bounded by the limits established in the ASME
Code. Mechanical testing has shown that the structural strength of
repair sleeves under normal, upset, emergency, and faulted
conditions provides margin to the acceptance limits. These
acceptance limits bound the most limiting (three times normal
operating pressure differential) burst margin recommended by NRC's
[U.S. Nuclear Regulatory Commission] Regulatory Guide 1.121, ``Bases
for Plugging Degraded PWR [Pressurized Water Reactor] Steam
Generator Tubes.'' Burst testing of sleeve/tube assemblies has
confirmed the analytical results and demonstrated that no
unacceptable levels of primary-to-secondary leakage are expected
during any plant condition.
The Alloy 800 repair sleeve depth-based structural limit is
determined using NRC guidance and the pressure stress equation of
ASME Code, Section III, with additional margin added to account for
configuration of long axial cracks. A bounding detection threshold
value has been conservatively identified and statistically
established to account for growth and determine the repair sleeve/
tube assembly plugging limit. A sleeved tube is plugged on detection
of degradation in the sleeve/tube assembly.
Evaluation of the repaired steam generator tube testing and
analysis indicates no detrimental effects on the sleeve or sleeved
tube assembly from reactor system flow, primary or secondary coolant
chemistries, thermal conditions or transients, or pressure
conditions as may be experienced at Comanche Peak Steam Electric
Station (CPSES), Unit 1 and Unit 2. Corrosion testing and historical
performance of sleeve/tube assemblies indicates no evidence of
sleeve or tube corrosion considered detrimental under anticipated
service conditions.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant or the manner in
which it is operated. The consequences of a hypothetical failure of
the sleeve/tube assembly is bounded by the current steam generator
tube rupture (SGTR) analysis described in CPSES Updated Final Safety
Analysis Report. Due to the slight reduction in the inside diameter
caused by the sleeve wall thickness, primary coolant release rates
would be slightly less than assumed for the steam generator tube
rupture analysis and, therefore, would result in a lower total
primary fluid mass release to the secondary system. A main steam
line break or feedwater line break will not cause a SGTR since the
sleeves are analyzed for a maximum accident differential pressure
greater than that predicted in the CPSES safety analysis. The
minimal repair sleeve/tube assembly leakage that could occur during
plant operation is well within the Technical Specification leakage
limits when grouped with current alternate plugging criteria
calculated leakage values.
Therefore, TXU Energy has concluded that the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The Alloy 800 leak limiting repair sleeves are designed using
the applicable ASME Code as guidance; therefore, it meets the
objectives of the original steam generator tubing. As a result, the
functions of the steam generators will not be significantly affected
by the installation of the proposed sleeve. The proposed repair
sleeves do not interact with any other plant systems. Any accident
as a result of potential tube or sleeve degradation in the repaired
portion of the tube is bounded by the existing SGTR accident
analysis. The continued integrity of the installed sleeve/tube
assembly is periodically verified by the Technical Specification
requirements and the sleeved tube will be plugged on detection of
degradation.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant, or the manner in
which it is operated. Therefore, TXU Energy concludes that this
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The repair of degraded steam generator tubes with Alloy 800 leak
limiting repair sleeves restores the structural integrity of the
degraded tube under normal operating and postulated accident
conditions and thereby maintains current core cooling margin as
opposed to plugging the tube and taking it out of service. The
design safety factors utilized for the repair sleeves are consistent
with the safety factors in the ASME Boiler and Pressure Vessel Code
used in the original steam generator design. The portions of the
installed sleeve/tube assembly that represent the reactor coolant
pressure boundary can be monitored for the initiation of sleeve/tube
wall degradation and the affected tube plugged on detection of
degradation. Use of the previously identified design criteria and
design verification testing assures that the margin of safety is not
significantly different from the original steam generator tubes.
Therefore, TXU Energy 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
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.
[[Page 49821]]
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 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].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: August 12, 2002, as supplemented
on February 28, 2003.
Brief description of amendment: The amendment changes the
surveillance requirements for the emergency diesel generators (EDGs) in
Technical Specification (TS) 3/4.8.1.1, ``Electrical Power Systems--
A.C. Sources--Operating'' and TS 3/4.8.1.2, ``Electrical Power
Systems--Shutdown.'' In addition, TS Section 6.0, ``Administrative
Controls,'' has been revised to add a new TS to define the program
requirements for testing the EDG fuel oil.
Date of issuance: July 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 277.
Facility Operating License No. DPR-65: This amendment revised the
TSs.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58639). The supplement dated February 28, 2003, provided additional
information which 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 July 25, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: August 14, 2002, as supplemented
on April 7, 2003.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) related to Containment Systems. Specifically, the
amendment: (1) Adds a new requirement for a Containment Tendon
Surveillance Program to TS Section 6.0, ``Administrative Controls;''
(2) deletes TS 3/4.6.1.6, ``Containment Structural Integrity;'' (3)
revises TS 3/4.6.1.1, ``Containment Integrity,'' to add a new
surveillance requirement that requires that containment structural
integrity be verified in accordance with the Containment Tendon
Surveillance Program; (4) revises TS 3/4.6.3.1, ``Containment Isolation
Valves,'' to add a new action statement that increases the allowed
outage time from 4 hours to 72 hours for Containment Isolation Valves
(CIVs) in closed systems; (5) makes other changes to the TSs for
Containment Integrity and CIVs to provide clarity to the TSs; and (6)
makes other administrative changes. In addition, the TS Bases have been
revised to address the proposed changes.
Date of issuance: July 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 278.
Facility Operating License No. DPR-65: This amendment revised the
TSs.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58640). The supplement dated April 7, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 25, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: August 12, 2002, as supplemented
on October 21, 2002, and January 15, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.8.2.3, ``Electrical Power Systems, D.C.
Distribution--Operating;'' TS 3.8.2.4, ``Electrical Power Systems, D.C.
Distribution--Shutdown;'' and TS 3.8.2.5, ``Electrical Power Systems,
D.C. Distribution Systems (Turbine Battery)--Operating'' to use
standard TS terminology in order to provide enhanced readability and
usability. The amendment also provides additional criteria for
determining battery operability upon restoration from a recharge or
equalizing charge.
Date of issuance: July 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 279.
Facility Operating License No. DPR-65: This amendment revises the
TSs.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61677). The supplements dated October 21, 2002, and January 15, 2003,
provided additional information which 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 July 29, 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: November 20, 2002, as
supplemented by letters dated January 21 and June 4, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications Required Actions requiring suspension of
operations involving positive reactivity additions and various Notes
that preclude reduction in boron concentration.
Date of issuance: July 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 207 & 201.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003, (68 FR
18273). The supplement dated June 4, 2003, provided clarifying
information that did not change the scope of the November 20, 2002,
application and its supplement dated January 21, 2003, nor
[[Page 49822]]
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 29, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket No. 50-370, McGuire Nuclear Station,
Unit 2, Mecklenburg County, North Carolina
Date of application for amendments: January 31, as supplemented by
letter dated May 1, 2003.
Brief description of amendments: The amendment authorizes a
revision to the Updated Final Safety Analysis Report to allow the
degassing and straightening of a bent Mark-BW irradiated fuel rod in
the McGuire spent fuel pool.
Date of issuance: August 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 198.
Facility Operating License Nos. NPF-17: Amendment authorized
revision of the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18274). The supplement dated May 1, 2003, provided clarifying
information that did not change the scope of the January 31, 2003,
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 August 4, 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: November 20, 2002, as
supplemented by letters dated January 21 and June 4, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications Required Actions requiring suspension of
operations involving positive reactivity additions and various Notes
that preclude reduction in boron concentration.
Date of issuance: July 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 216 & 197.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18273). The supplement dated June 4, 2003, provided clarifying
information that did not change the scope of the November 20, 2002,
application and its supplement dated January 21, 2003, nor the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 29, 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 24, 2002, as supplemented
on June 23, 2003.
Brief description of amendment: The amendment revises Technical
Specification Surveillance Requirement (SR) 3.7.7.2 to require all city
water header isolation valves be open rather than only the one header
supply isolation valve. On June 23, 2003, the licensee withdrew its
request for changes to SR 3.7.7.1 pertaining to the city water tank
volume.
Date of issuance: August 4, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 218.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50952). The June 23, 2003, letter provided clarifying information that
did not enlarge the scope of the original Federal Register notice 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 August 4, 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 amendment: January 31, 2003.
Brief description of amendment: The amendment revises the Technical
Specification allowable values for two isolation condenser system
isolation functions, namely the Steam Flow--High and Return Flow--High,
for Units 2 and 3.
Date of issuance: July 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 200.
Facility Operating License No. DPR-19: The amendment revised the
Technical Specifications.
Amendment No.: 192.
Facility Operating License No. DPR-25: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 30, 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: January 27, 2003.
Brief description of amendment: The amendment revised Technical
Specifications (TS) 2.1.6, 3.2 (Table 3-5), and 5.9.1c as follows:
(1) TS 2.1.6(1), the ``as-found'' pressurizer safety valve (PSV)
lift setting tolerance band of +/-1% is increased to +1%/-3% to allow
for normal setpoint variance for Modes 1 and 2. The Basis of TS 2.1.6
is revised to clarify that the PSVs are still operable and capable of
performing their safety function with the wider tolerance band. The
other revisions to TS 2.1.6 are administrative in nature to change
defined terms to upper case text.
(2) TS 3.2, Table 3-5, Item 3 is revised to require an ``as-left''
PSV lift setting tolerance band of +/-1%.
(3) TS 5.9.1c is revised to remove the requirement to provide a
statement in the Monthly Operating Report (MOR) concerning failures or
challenges to power operated relief valves or safety valves. Generic
Letter 97-02, ``Revised Contents of the Monthly Operating Report,''
does not require the MOR to provide this information.
Date of issuance: July 25, 2003.
Effective date: July 25, 2003, and shall be implemented within 30
days from the date of issuance.
Amendment No.: 129.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12956).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 25, 2003.
No significant hazards consideration comments received: No.
[[Page 49823]]
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: May 23, 2002.
Brief description of amendment: The proposed amendment revises the
Unit 2 Operating License and several sections of Technical
Specifications to delete information differentiating between Unit 1 and
Unit 2 specific to Model E steam generators.
Date of issuance: July 21, 2003.
Effective date: As of the date of issuance and shall be implemented
120 days from the date of issuance.
Amendment Nos.: Unit 1-154; Unit 2-142.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42831). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 21, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant
(SQN), Unit 2, Hamilton County, Tennessee
Date of application for amendment: June 5, 2003.
Description of amendment: The amendment revised the reactor coolant
system heatup and cooldown curves (pressure-temperature (P-T) limits).
The revision replaced the P-T limits that were analyzed for 14.5
Effective Full Power Years (EFPYs) with new limits analyzed for 32
EFPYs. In addition, the amendment included corresponding changes to the
Technical Specification (TS) figure associated with the Low Temperature
Over Pressure Protection and the TS Bases.
Date of issuance: July 31, 2003.
Effective date: As of the date of issuance and shall be implemented
within 15 days of issuance.
Amendment No.: 277.
Facility Operating License No. DPR-79: Amendment revises the TSs.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37583).
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible from the Agencywide Documents Assess 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 Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or
by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By September 18, 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
[[Page 49824]]
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,
and electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in
accessing the document, contact the PDR Reference staff at 1-800-397-
4209, 301-415-4737, or by e-mail to [email protected]. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings 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 the continuing disruptions in delivery of mail to United
States Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the petition for
leave to intervene and request for hearing 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One (ANO), Units 1 and 2, Pope County, Arkansas
Date of amendment request: February 24, 2003, as supplemented by
letters dated March 25, June 30, and July 21, 2003.
Description of amendment request: The amendments allow the licensee
to use the spent fuel crane (L-3 crane) to lift heavy loads in excess
of 100 tons. Specifically the licensee received approval to use the
upgraded L-3 crane for loads up to 130 tons.
Date of issuance: July 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 220/248.
Facility Operating License Nos. (DPR-51 and NPF-6): Amendments
allow use of the upgraded L-3 crane to lift loads up to 130 tons.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (68 FR 11157, dated March 7, 2003, and 68 FR
41020, dated July 9, 2003). The notices provided an opportunity to
submit comments on the Commission's proposed NSHC determination. No
comments have been received. The notices also provided an opportunity
to request a hearing by April 7, 2003, and July 23, 2003, but indicated
that if the Commission makes a final NSHC determination, any such
hearing would take place after issuance of the amendment.
The July 21, 2003, supplemental letter provided clarifying
information that did not change the scope of the Federal Register
notice or the NSHC determination published July 9, 2003 (68 FR 41020).
The Commission's related evaluation of the amendments, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a Safety Evaluation dated July 25, 2003.
[[Page 49825]]
Attorney for licensee: Winston and Strawn, 1400 L Street, NW.,
Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Dated at Rockville, Maryland, this 11th day of August 2003.
For The Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-20839 Filed 8-18-03; 8:45 am]
BILLING CODE 7590-01-P