[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