[Federal Register Volume 68, Number 42 (Tuesday, March 4, 2003)]
[Notices]
[Pages 10277-10286]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-4623]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, February 7, 2003, through February 20,
2003. The last biweekly notice was published on February 18, 2003 (68
FR 7810).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By April 3, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714,\1\ which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29,
2002.
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As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
[[Page 10278]]
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, 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].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: January 14, 2003.
Description of amendment request: The proposed amendment would
revise the TMI-1 Technical Specification Sections 3.8.9, 3.15.2, and
4.12.2, and the associated Bases to delete the requirements for the
Reactor Building Purge Air Treatment System.
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.
This change will delete the existing Technical Specifications
3.15.2 and 4.12.2 and revise Technical Specification 3.8.9. The
proposed change does not impact nor change the physical
configuration of any system, structure or component, nor does it
change the manner in which any system is operated. Any change to the
system design will be evaluated in accordance with the requirements
of 10 CFR 50.59. Failure of the system will neither initiate any
type of accident nor increase the severity of the consequences of an
accident previously evaluated. Previously approved analyses of the
dose consequences of the accidents described in the TMI Unit 1 UFSAR
[Updated Final Safety Analysis Report] are not affected by the
proposed change and dose consequences remain below the limits of 10
CFR 50.67 without the operation of the Reactor Building Purge Air
Treatment System fan and filter components. The Reactor Building
Purge Air Treatment System fan and filter components are not
required for mitigation of any accident as described in the TMI Unit
1 UFSAR. Reactor Building purge operations will continue to be
conducted in accordance with the existing plant administrative
controls, which will ensure the limits of 10 CFR part 50 Appendix I
are met.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 10279]]
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This activity will delete sections of the Technical
Specifications applicable to the Reactor Building Purge Air
Treatment System fan and filter components. The proposed change does
not physically alter any system, structure or component. Any change
to the system design will be evaluated in accordance with 10 CFR
50.59. The proposed change will not cause the Reactor Building Purge
Air Treatment System to operate outside of its existing design
basis. There will be no impact to any operational feature of the
system or any procedures that control its operation that could
result in a new or different failure mode. The design basis of the
Reactor Building Purge Air Treatment System as currently described
in the TMI Unit 1 UFSAR is not revised.
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 deletion of Technical Specification Sections 3.15.2 and
4.12.2 and the revision of Technical Specification 3.8.9 will not
impact the operation of the Reactor Building Purge Air Treatment
System. The proposed change will not cause the system to be placed
in a configuration outside of its design basis. The proposed change
will not reduce the margin of safety of any safety related system.
Reactor Building purge operations will continue to be conducted in
accordance with existing plant administrative controls, which will
ensure the limits of 10 CFR part 50 appendix I are met. The system
will continue to be operable in accordance with applicable plant
operating procedures.
The system will also continue to be tested and maintained under
periodic operations surveillance and the TMI Unit 1 Preventive
Maintenance Program.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice
President, General Counsel and Secretary, Exelon Generation Company,
LLC, 300 Exelon Way, Kennett Square, PA 19348.
NRC Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear
Power Station, Unit 2, Grundy County, Illinois
Date of amendment request: January 31, 2003.
Description of amendment request: The proposed amendments would
revise the safety limit minimum critical power ratio for Unit 2 for two
loop operation and for single loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC
[Nuclear Regulatory Commission] approved methods to ensure that fuel
performance during normal, transient, and accident conditions is
acceptable. The proposed change conservatively establishes the
safety limit for the minimum critical power ratio (SLMCPR) for
Dresden Nuclear Power Station (DNPS), Unit 2, Cycle 18 such that the
fuel is protected during normal operation and during any plant
transients or anticipated operational occurrences (AOOs).
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The proposed change revises the SLMCPR to protect the fuel
during normal operation as well as during any transients or
anticipated operational occurrences. Operational limits will be
established based on the proposed SLMCPR to ensure that the SLMCPR
is not violated during all modes of operation. This will ensure that
the fuel design safety criteria (i.e., that at least 99.9% of the
fuel rods do not experience transition boiling during normal
operation and anticipated operational occurrences) is met. Since the
operability of plant systems designed to mitigate any consequences
of accidents has not changed, the consequences of an accident
previously evaluated are not expected to increase.
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?
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. The proposed change does not involve
any modifications of the plant configuration or allowable modes of
operation. The proposed change to the SLMCPR assures that safety
criteria are maintained for DNPS, Unit 2, Cycle 18. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The value of the proposed SLMCPR provides a margin of safety by
ensuring that no more than 0.1% of the rods are expected to be in
boiling transition if the MCPR limit is not violated. The proposed
change will ensure the appropriate level of fuel protection.
Additionally, operational limits will be established based on the
proposed SLMCPR to ensure that the SLMCPR is not violated during all
modes of operation. This will ensure that the fuel design safety
criteria (i.e., that at least 99.9% of the fuel rods do not
experience transition boiling during normal operation as well as
AOOs) are met.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: December 20, 2002.
Description of amendment request: The proposed amendment would make
an administrative change to Technical Specification (TS) Sections 6.7,
6.14, and 6.15 by replacing ``Station Review Board'' to ``Plant
Operations Review Committee'' to be consistent with the name for this
type of onsite review committee that is used at other FirstEnergy
Nuclear Operating Company plants. Additionally, the proposed amendment
would make an administrative change to TS 6.8 to update the version of
Regulatory Guide 1.33 referenced in that Section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 10280]]
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The administrative changes do not affect any existing limits,
and accident initial conditions, probability, and assumptions remain
as previously analyzed. The proposed change to the name of the
onsite review committee or the version of the Regulatory Guide will
have no significant effect on accident initiation frequency. The
proposed changes do not invalidate the assumptions used in
evaluating the radiological consequences of any accident. Therefore,
the proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are administrative and do not introduce any
new or different accident initiators. Therefore, the proposed
changes do 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 proposed changes are administrative and will not have a
significant effect on any margin of safety. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
Based on the above, FENOC concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: January 14, 2003.
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TSs) for the control room
emergency ventilation system (CREVS) such that movement of irradiated
fuel assemblies will be allowed to commence with one CREVS
pressurization train inoperable, provided the appropriate TS Action
requirements are implemented.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No
Probability of Occurrence of an Accident Previously Evaluated
[Cook Nuclear Plant] CNP TS 3.0.4 requires that TS limiting
conditions for operation be met without reliance on the Action
statements prior to entering an Applicability condition. The
proposed change to the CNP CREVS TS to allow an exception to TS
3.0.4 during movement of irradiated fuel assemblies does not affect
any accident initiators or precursors. The CREVS function is purely
mitigative. There is no design basis accident that is initiated by a
failure of the CREVS function. An exception to TS 3.0.4 will not
create any adverse interactions with other systems that could result
in initiation of a design basis accident. Therefore, the probability
of occurrence of an accident previously evaluated is not
significantly increased.
Consequences of an Accident Previously Evaluated
The accident consequence that is relevant to the proposed change
is the dose to control room personnel from a fuel handling accident.
The CNP licensing basis analysis of a fuel handling accident has
determined that the dose would be within the applicable limits of
GDC 19. The current TS specify actions to be taken if one CREVS
pressurization train is inoperable during movement of irradiated
fuel assemblies. These actions provide assurance that the CREVS will
perform its mitigating function as assumed in the accident analysis.
Since the proposed change will continue to require these actions,
the fuel handling accident analysis will remain valid. Therefore,
the consequences of an accident previously analyzed are not
significantly increased.
In summary, the probability of occurrence and the consequences
of an accident previously evaluated are not significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No
The proposed change does not create any new or different
accident initiators or precursors. The option to commence movement
of irradiated fuel assemblies while relying on the provisions of the
Action statement does not affect the manner in which any accident
begins. The proposed change does not create any new accident
scenarios and does not change the interaction between the CREVS and
any other system. Thus, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No
The margin of safety associated with the proposed change is that
associated with the applicable control room dose limit specified by
GDC 19. The proposed change will continue to require actions that
assure the dose to control room personnel determined by the fuel
handling accident analysis remains valid. Therefore, the proposed
change does not involve a significant reduction in margin of safety.
In summary, based upon the above evaluation, [Indiana Michigan
Power] I&M has concluded that the proposed change involves 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: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of amendment request: January 14, 2003.
Description of amendment request: The proposed one-time change
revises the steam generator inservice inspection frequency requirements
in Technical Specification 4.4.5.3.a for V.C. Summer Nuclear Station
(VCSNS) immediately after refueling outage RF-12. The change would
allow a 58-month maximum inspection interval after two inspections
resulting in C-1 classification, rather than a 40-month maximum
inspection interval. This change is proposed to eliminate premature/
unnecessary steam generator inspections, due to a shortened operating
cycle, which will result in significant dose and schedule impacts.
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
[[Page 10281]]
consequences of an accident previously evaluated?
Response: No.
The proposed one-time extension of the Technical Specification
inspection interval does not involve changing any structure, system
or component or affect plant operations. It is not an initiator of
any accident and does not change any FSAR [Final Safety Analysis
Report] safety analyses. As such, the proposed change does not
involve a significant increase in the probability of an accident
previously evaluated.
Probability of an Accident
The VCSNS Steam Generator Management Program includes provisions
that are more rigorous than existing Technical Specification
requirements. The topics addressed by the program include:
[sbull] Steam generator performance criteria, including a
reduced operational leakage limit.
[sbull] Steam generator repair criteria and repair methods.
[sbull] Steam generator inspections that include Degradation
Assessments, Condition Monitoring Assessments, and Operational
Assessments.
[sbull] NDE [nondestructive examination] technique requirements.
The results of the above program requirements demonstrated that
all performance requirements were met during Refuel 12.
Consequences of an Accident
The consequences of design basis accidents are, in part,
functions of the specific activity in the primary coolant and the
primary to secondary leakage rates resulting from an accident.
Therefore, limits are included in the Technical Specifications for
operational leakage and for specific activity in the reactor coolant
to ensure the plant is operated in its analyzed condition.
The VCSNS program requires a 150-gallon per day per steam
generator limit for leakage prior to an accident. This limit is a
reduction in the current Technical Specification value. The post
accident leak rate remains at the same value assumed by the accident
analysis (1 gallon per minute). Since the new operational leakage
limit is more conservative than the existing value, it will not
increase the likelihood or consequences of an accident.
In consideration of the above, past 100% eddy current results
after 5.4 EFPY [effective full-power years] of operation, and the
current leak free condition of the steam generators, extending the
tube inspection frequency does not involve a significant increase in
the consequences of a previously evaluated accident.
Summary
The proposed change does not affect the design of the steam
generators, their method of operation, or primary coolant chemistry
controls. The change does not adversely impact any other previously
evaluated design basis accident.
Therefore, the change does not affect the consequences of a SGTR
[steam generator tube rupture] or any other accident.
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 one-time extension of the Technical Specification
inspection interval does not involve changing any structure, system
or component or affect plant operations. It is not an initiator of
any accident and does not change any FSAR safety analyses.
Primary to secondary leakage that may be experienced during
plant conditions is expected to remain within current accident
analysis assumptions.
The proposed change does not affect the design of the steam
generators, their method of operation, or primary coolant chemistry
controls. In addition, the change does not impact any other plant
system or component.
Therefore, the change does not create the possibility of a new
or different type of accident or malfunction from any accident
previously evaluated.
3. Does this change involve a significant reduction in margin of
safety?
Response: No.
The steam generator tubes are an integral part of the reactor
coolant pressure boundary and, as such, are relied upon to maintain
the primary system pressure and inventory. As part of the RCS
[reactor coolant system] boundary, the tubes are unique in that they
are also relied upon as a heat transfer medium between the primary
and secondary systems such that heat may be removed from the primary
system. Additionally, the steam generator tubes also isolate the
radioactive fission products in the primary coolant from the
secondary system. In summary, the safety function of the steam
generator is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. Extending the
tube inspection frequency will not alter the design function of the
steam generators. Previous inspections conducted during Refuel 12
demonstrate that there is no active tube damage mechanism. The
improved design of the Model Delta 75 generator also provides
reasonable assurance that leakage is not likely to occur over the
next operating period.
For the above reasons, the margin of safety is unchanged and
overall plant safety will be maintained by the proposed Technical
Specification revision.
Pursuant to 10 CFR 50.91, the preceding analyses provide a
determination that the proposed Technical Specification change poses
no significant hazard as delineated by 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas G. Eppink, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: John A. Nakoski.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: January 14, 2003.
Description of amendment request: The proposed change will exclude
the Charging/Safety Injection (SI) pumps and the Residual Heat Removal
pumps from the requirement to vent emergency core cooling system pump
casings located in Technical Specification (TS) Section 4.5.2.b.2,
eliminate the 31-day venting surveillance for the SI pumps, and add
discussion for this exclusion in the Technical Basis of TS Section B 3/
4.5.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to Technical Specification 4.5.2.b.2 and
its associated bases do not contribute to the initiation of any
accident previously evaluated. Supporting factors are as follows:
[sbull] The safety function of the Charging/SI system, which is
related to accident mitigation, has not been altered. Therefore, the
probability of an accident is not increased by the exclusion of the
Charging/SI system discharge venting requirements.
[sbull] The exclusion of the Charging/SI system venting
requirements does not affect the integrity of the Charging/SI system
such that its function in the control of radiological consequences
is affected. In addition, the exclusion of the Charging/SI system
venting requirements does not alter any fission product barrier. The
exclusion of the Charging/SI system venting requirements does not
change, degrade, or prevent the response of the Charging/SI system
to accident scenarios, as described in FSAR [Final Safety Analysis
Report] Chapter 15. In addition, the exclusion of the Charging/SI
system venting requirements does not alter any assumptions
previously made in the radiological consequence evaluations nor
affect the mitigation of the radiological consequences of an
accident described in the FSAR. Therefore, the consequences of an
accident previously evaluated in the FSAR will not be increased.
[sbull] The clarification of the RHR [residual heat removal]
pump piping venting does not affect the integrity of the RHR system
such that its function in the control of radiological consequences
is affected. In addition, the
[[Page 10282]]
clarification does not alter any of the fission product barriers.
The clarification does not change, degrade, or prevent the response
of the RHR system to accident scenarios, as described in FSAR
Chapter 15. In addition, the clarification to the RHR pump piping
venting does not alter any assumption previously made in the
radiological consequences evaluations nor affect the mitigation of
the radiological consequences of an accident described in the FSAR.
Therefore, the consequences of an accident previously evaluated in
the FSAR will not be increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to Technical Specification 4.5.3.b.2 and
its associated bases do not introduce any new accident initiator
mechanisms. The clarification of the RHR pump piping venting and the
exclusion of the Charging/SI system venting requirements does not
cause the initiation of any accident nor create any new credible
limiting single failure. The exclusion of the Charging/SI system
venting requirements does not result in any event previously deemed
incredible being made credible. As such, it does not create the
possibility of an accident different than any evaluated in the FSAR.
3. Does this change involve a significant reduction in margin of
safety?
Response: No.
The exclusion of the Charging/SI system venting requirements
does not result in a condition where the design, material, and
construction standards that were acceptable prior to this change of
the Charging/SI or RHR system venting requirements are altered. The
proposed changes to Technical Specification 4.5.2.b.2 and its
associated bases will have no affect on the availability,
operability, or performance of the Charging/SI or RHR systems.
Therefore, the clarification of the RHR pump piping venting and the
exclusion of the Charging/SI system venting requirements will not
reduce the margin of safety, as described in the bases to any
technical specification.
Pursuant to 10 CFR 50.91, the preceding analyses provide a
determination that the proposed Technical Specifications change
poses no significant hazard as delineated by 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas G. Eppink, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: John A. Nakoski.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: July 25, 2002 as supplemented by letter
dated February 5, 2003.
Brief description of amendments: The proposed amendments would
change the CPSES Facility Operating Licenses as follows: Section
2.C.(4)(b) would be changed to be consistent with the license
conditions stated in the U.S. Nuclear Regulatory Commission (NRC) Order
and Safety Evaluation issued December 21, 2001, which approved the
direct transfer of ownership interest and operating authority for CPSES
to TXU Generation Company LP; Section 2.E, which requires reporting any
violations of the requirements contained in Section 2.C of the
licenses, would be deleted. Additionally, Technical Specification Table
5.5-2 ``Steam Generator Tube Inspection,'' Table 5.5-3, ``Steam
Generator Repaired Tube Inspection for Unit 1 Only,'' and Section
5.6.10, ``Steam Generator Tube Inspection Report,'' would be revised to
delete the requirement to notify the NRC pursuant to Section
50.72(b)(2), ``Immediate notification requirements for operating
nuclear power reactors,'' of Title 10 of the Code of Federal
Regulations (10 CFR) if the steam generator tube inspection results are
in a C-3 classification. The basis for the proposed no significant
hazards consideration determination associated with the application was
published in the Federal Register on September 3, 2002 (67 FR 56329).
By letter dated February 5, 2003, TXU Generation Company, LP
requested that the proposed change to license conditions in Section
2.C.(4)(b) be superseded by the proposed deletion of the license
conditions, related to Decommissioning Trusts, specified in Sections
2.C.(4)(a), 2.C.(4)(b), 2.C.(4)(d), 2.C.(4)(e), and 2.C.(6).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), ``Notice for public
comment; State consultation,'' the licensee has provided its analysis
of the issue of no significant hazards consideration, as they relate to
the February 5, 2003 supplement, 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 requested changes delete certain license conditions
pertaining to Decommissioning Trust Agreements currently in Sections
2.C.(4)(a), 2.C.(4)(b), 2.C.(4)(d), 2.C.(4)(e), and 2.C.(6) of the
CPSES Facility Operating Licenses (NPF-87 and NPF-89). The requested
changes are consistent with the NRC's Final Rule for Decommissioning
Trust Provisions as published in the Federal Register on December
24, 2002 (67 FR 78332).
The revised regulations of 10 CFR 50.75(h)(4)[, ``Reporting and
recordkeeping for decommissioning planning,''] state ``Unless
otherwise determined by the Commission with regard to a specific
application, the Commission has determined that any amendment to the
license of a utilization facility that does no more than delete
specific license conditions relating to the terms and conditions of
decommissioning trust agreements involves ``no significant hazard[s]
consideration'.''
This request involves administrative changes only. No actual
plant equipment or accident analyses will be affected by the
proposed changes. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This request involves administrative changes only to be
consistent with the NRC's Final Rule for Decommissioning Trust
Provisions as published in the Federal Register (67 FR 78332).
No actual plant equipment or accident analyses will be affected
by the proposed change and no failure modes not bounded by
previously evaluated accidents will be created. Therefore, the
proposed changes do not create a new or different kind of accident
from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
This request involves administrative changes only to be
consistent with the NRC's Final Rule for Decommissioning Trust
Provisions as published in the Federal Register (67 FR 78332).
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel and fuel cladding, Reactor
Coolant System pressure boundary, and containment structure) to
limit the level of radiation dose to the public.
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed changes will not
relax any criteria used to establish safety limits, will not relax
any safety systems settings, or will not relax the bases for any
limiting conditions of operation. Therefore, the proposed 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),
``Issuance of amendment,'' are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
[[Page 10283]]
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.
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 email to [email protected].
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 application for amendments: November 13, 2002, as
supplemented November 20, 2002.
Brief description of amendments: The amendments delete Technical
Specification 5.5.3, ``Post Accident Sampling System (PASS),'' and
thereby eliminate the requirements to have and maintain the PASS at
Brunswick Steam Electric Plant, Units 1 and 2.
Date of issuance: February 11, 2003.
Effective date: February 11, 2003, to be implemented within 180
days of issuance.
Amendment Nos.: 226 & 253.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
799).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 11, 2003.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 14, 2002, as supplemented by letter
dated December 6, 2002.
Brief description of amendment: The amendment revises the technical
specification safety function lift setpoint tolerances for the Safety/
Relief valves (S/RVs). The changes also allow surveillance of the
relief mode of operation of the S/RVs to be performed without
physically lifting the disk of a valve off the seat at power.
Date of issuance: February 13, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 130.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42822).
The December 6, 2002, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 13, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of application for amendments: April 19, 2002, as supplemented
by letters dated September 9, 2002 and January 3, 2003.
Brief description of amendments: The amendments revise Technical
Specifications (TS) 3.6.6, ``Containment Spray and Cooling Systems,''
to change the frequency of Surveillance Requirement (SR) 3.6.6.8 from
``10 years'' to ``Following maintenance that could result in nozzle
blockage OR Following fluid flow through nozzles.''
Date of issuance: February 20, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 126.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 11, 2002 (67 FR
40023) The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The safety evaluation addresses Braidwood Station Units 1 and 2
only. The NRC staff's evaluation of the Byron Units 1 and 2 will be
addressed separately.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 20, 2003.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: June 5, 2002, as supplemented
August 13, September 30, October 31, November 13, and November 25,
2002.
Brief description of amendment: The amendment approves an increase
in maximum steady-state core power level from 2544 megawatts thermal
(MWt) to 2568 MWt, an increase of approximately 0.9 percent.
Date of issuance: December 4, 2002.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 205.
Facility Operating License No. DPR-72: Amendment revises the
Facility
[[Page 10284]]
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42826). The August 13, September 30, October 31, November 13, and
November 25, 2002, supplements contained clarifying information only
and did not change the initial no significant hazards consideration
determination or expand the scope of the initial application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 4, 2002.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: June 13, 2002.
Brief description of amendment: The amendment revises Improved
Technical Specification (ITS) 3.3.8, ``Emergency Diesel Generator (EDG)
Loss of Power Start (LOPS),'' by changing the completion time for
required action D.2 from 12 to 36 hours. The amendment also corrects a
typographical error in ITS 3.3.8.
Date of issuance: February 11, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 206.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45570).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 11, 2003.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County,
California
Date of application for amendments: November 16, 2001, as
supplemented by letter dated September 13, 2002.
Brief description of amendments: The amendments revise Technical
Specification (TS) 1.1, ``Definitions, Dose Equivalent I-131,'' and
authorize revision of the Final Safety Analysis Report (FSAR) Update to
reflect the revised steam generator tube rupture and main steam line
break radiological consequences analyses.
Date of issuance: February 20, 2003.
Effective date: February 20, 2003, and shall be implemented in the
next periodic update to the FSAR Update.
Amendment Nos.: Unit 1--156; Unit 2--156.
Facility Operating License Nos. DPR-80 and DPR-82: The amendment
revised the Technical Specifications and the FSAR Update.
Date of initial notice in Federal Register: January 8, 2002 (67 FR
931). The September 13, 2002, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 20, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: March 4, 2002 (TS 01-03).
Brief description of amendments: The amendments revise the SQN Unit
1 and 2 Technical Specifications (TSs) by deleting one definition and
modifying several subsections contained in TS Section 6.0,
``Administrative Controls.'' These changes have been prepared based on
existing NRC guidance.
Date of issuance: February 11, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: 281 & 272.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TSs.
Date of initial notice in Federal Register: April 16, 2002 (67 FR
18649). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 11, 2003.
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.
[[Page 10285]]
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 April 3, 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,\2\ 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.
---------------------------------------------------------------------------
\2\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and paragraph (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April
29, 2002.
---------------------------------------------------------------------------
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
[[Page 10286]]
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).
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: January 16, 2003, as
supplemented on January 31, 2003.
Brief description of amendment: The amendment modifies Technical
Specification 3.1.7 to permit the use of an alternate method of
determining rod position for Control Rod H-10 until the end of Cycle 22
or until the next shutdown of sufficient duration, whichever occurs
first.
Date of issuance: February 13, 2003.
Effective date: February 13, 2003.
Amendment No. 197.
Facility Operating License No. DPR-23. Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (68 FR 3556 dated January 24, 2003). The
notice provided an opportunity to submit comments on the Commission's
proposed NSHC determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by February 24,
2003, but indicated that if the Commission makes a final NSHC
determination, any such hearing would take place after issuance of the
amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final determination of NSHC are contained in
a Safety Evaluation dated February 13, 2003.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen G. Howe.
Dated at Rockville, Maryland, this 21st day of February 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-4623 Filed 3-3-03; 8:45 am]
BILLING CODE 7590-01-P