[Federal Register Volume 67, Number 228 (Tuesday, November 26, 2002)]
[Notices]
[Pages 70762-70775]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-29737]


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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 November 1, 2002, through November 14, 2002. 
The last biweekly notice was published on November 12, 2002 (67 FR 
68727).

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, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By December 26, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to

[[Page 70763]]

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, 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 
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, 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, 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, 304-415-4737 or by e-mail to [email protected].

Arizona Public Service Company, et al., Docket No. STN 50-528, Palo 
Verde Nuclear Generating Station, Unit 1, Maricopa County, Arizona

    Date of amendment request: September 26, 2002, as supplemented by 
letter dated October 23, 2002.

[[Page 70764]]

    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.5.9, ``Steam Generator (SG) Tube 
Surveillance Program,'' to clearly delineate the scope of the tube 
inspection required in the SG tubesheet region. TS 5.5.9 is in section 
5, ``Administration Controls,'' of the TSs.
    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.
    Arizona Public Service Company (APS) proposes to modify Palo 
Verde Nuclear Generating Station (PVNGS) Technical Specifications 
for Unit 1 to define the SG tube inspection scope. The PVNGS Unit 1 
specific analysis takes into account the reinforcing effect the 
tubesheet has on the external surface of an expanded SG tube. Tube-
bundle integrity will not be adversely affected by the 
implementation of the revised tube inspection scope. SG tube burst 
or collapse cannot occur within the confines of the tubesheet; 
therefore, the tube burst and collapse criteria of NRC Regulatory 
Guide (RG) 1.121 (Bases for Plugging Degraded PWR Steam Generator 
Tubes) are inherently met. Any degradation below the TEA (Tube 
Engagement Area) length is shown by analyses and test results to be 
acceptable, thereby precluding an event with consequences similar to 
a postulated tube rupture event.
    Tube burst is precluded for cracks within the tubesheet by the 
constraint provided by the tubesheet. Thus, structural integrity is 
maintained by the tubesheet constraint. However, a 360-degree 
circumferential crack or many axially oriented cracks could permit 
severing of the tube and tube pullout from the tubesheet under the 
axial forces on the tube from primary to secondary pressure 
differentials. Testing was performed to define the length of non-
degraded tubing that is sufficient to compensate for the axial 
forces on the tube and thus prevent pullout. This proposed amendment 
would encompass that length of non-degraded tubing for inspection.
    In conclusion, incorporation of the revised inspection scope 
into PVNGS Unit 1 Technical Specifications maintains existing design 
limits and therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Tube-bundle integrity is expected to be maintained during all 
plant conditions upon implementation of the proposed tube inspection 
scope. Use of this scope does not introduce a new mechanism that 
would result in a different kind of accident from those previously 
analyzed. Even with the limiting circumstances of a complete 
circumferential separation of a tube occurring below the TEA length, 
SG tube pullout is precluded and leakage is predicted to be 
maintained within the Updated Final Safety Analysis Report limits 
during all plant conditions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Upon implementation of the revised inspection scope, operation 
with potential cracking below the Inspection Extent length in the 
explansion region of the SG tubing meets the margin of safety as 
defined by RG 1.121 and RG 1.83 (Inservice Inspection of Pressurized 
Water Reactor Steam Generator Tubes) and the requirements of General 
Design Criteria 14, 15, 31, and 32 of 10 CFR (part) 50. Accordingly, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    Based on the above evaluation, APS 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 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendment involves no significant hazards consideration.
    The above amendment was previously noticed in the Federal Register 
on October 3, 2002 (67 FR 62079), as an exigent circumstances TS 
amendment, based on the preliminary determination that the TS amendment 
was needed on or about October 25, 2002, to allow Unit 1 to restart 
from its refueling outage. On further consideration, it has been 
determined that the proposed TS amendment does not have to be issued 
before the restart of Unit 1. This notice supersedes and replaces the 
exigent circumstances TS amendment notice of October 3, 2002.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: September 26, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.9.1, ``Refueling Equipment 
Interlocks,'' to allow fuel movement to continue if the refueling 
interlocks become inoperable, and add two new alternative Required 
Actions for the condition when the refueling equipment interlocks are 
inoperable. Specifically, the proposed amendment would add Required 
Actions 3.9.1.A.2.1 to immediately block control rod withdrawal and 
3.9.1.A.2.2 to perform a verification that all of the control rods are 
fully inserted. The proposed changes are similar to the proposed 
generic change that was provided in Technical Specifications Task Force 
(TSTF) Traveler, TSTF-225, revision 1, ``Fuel Movement With Inoperable 
Refueling Equipment Interlocks,'' dated November 22, 2000, for the NRC 
staff's review.
    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 amendment to the Technical Specifications does not 
result in the alteration of the design, material, or construction 
standards that were applicable prior to the change. The same 
Refueling Interlocks instrumentation is used, and the control rod 
removal error and fuel assembly insertion error assumptions in the 
Updated Final Safety Analysis Report (UFSAR) chapter 15 analysis 
remain unchanged. The proposed additional Required Actions provide 
an equivalent level of assurance that fuel will not be loaded into a 
core cell with a control rod withdrawn as does the current TS 
Required Action. The proposed change will not result in the 
modification of any system interface that would increase the 
likelihood of an accident since these events are independent of the 
proposed change. The proposed amendment will not change, degrade, or 
prevent actions, or alter any assumptions previously made in 
evaluating the radiological consequences of an accident described in 
the UFSAR. Therefore, the proposed amendment does not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change in the TS requirements does not alter the 
performance of the Refueling Equipment Interlocks. The change does 
not involve a change in plant design or to the analyzed condition of 
the reactor core during

[[Page 70765]]

refueling. The proposed new Required Actions will ensure that 
control rods are not withdrawn and cannot be inappropriately 
withdrawn because a block to control rod withdrawal is in place. 
Implementation of the proposed amendment does not create the 
possibility of a new of different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    As discussed in the Bases for the affected TS requirements, 
inadvertent criticality is prevented during the loading of fuel 
provided all control rods are fully inserted. The refueling 
interlocks function to support the refueling procedures by 
preventing control rod withdrawal during fuel movement, and the 
inadvertent loading of fuel when a control rod is withdrawn. The 
proposed change will allow the refueling interlocks to be inoperable 
and fuel movement to continue, only if a control rod withdrawal 
block is in effect and all control rods are verified to be fully 
inserted. These proposed Required Actions provide an equivalent 
level of protection as the refueling interlocks by preventing a 
configuration which could lead to an inadvertent criticality event. 
The refueling procedures will continue to be supported by the 
proposed Required Actions because control rods cannot be withdrawn 
and as a result, fuel cannot be inadvertently loaded when a control 
rod is withdrawn. Plant and system response to an initiating event 
will remain in compliance within the assumptions of the safety 
analyses, and therefore, the margin of safety is not affected. 
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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: September 26, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement (SR) 
3.7.3.6 associated with the verification of the control room emergency 
filtration (CREF) system duct work unfiltered in-leakage. Specifically, 
the proposed amendment would add a note to SR 3.7.3.6 to allow 
crediting the performance of an integrated tracer gas test of the 
control room envelope while in the recirculation mode to satisfy the 
requirements of the surveillance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This license amendment proposes an alternative test for 
performing the CREF system surveillance associated with measuring 
the Control Room Envelope (CRE) unfiltered inleakage. The CREF 
system provides a configuration for mitigating radiological 
consequences of accidents; however, it does not involve the 
initiation of any previously analyzed accident. Therefore, the 
proposed change cannot increase the probability of any previously 
evaluated accident.
    The CREF system provides a radiologically controlled environment 
from which the plant can be safely operated following a radiological 
accident. Design basis accident analyses conclude that radiological 
consequences are within the regulatory acceptance criteria. The 
current Technical Specifications (TS) surveillance (SR 3.7.3.6) 
measures inleakage from four sections of CREF system duct work 
outside the CRE that are at negative pressure during accident 
conditions. The proposed Tracer Gas test provides a measurement of 
CRE inleakage from all potential sources including the four sections 
of duct work. The use of Tracer Gas testing in accordance with the 
methods described in American Society of Testing and Materials 
(ASTM) standard E741 has been accepted by both the NRC and the 
industry. Measuring the CRE inleakage using Tracer Gas testing has 
no effect on the CREF system function. The results of Tracer Gas 
testing will be assessed in accordance with regulatory guidance and 
industry guidance and compliance with 10 CFR [part] 50, Appendix A, 
General Design Criterion (GDC)-19 will be demonstrated. Therefore, 
the proposed change does not significantly increase the radiological 
consequences of any previously evaluated accident. Based on the 
above, the proposed change does not significantly increase the 
probability or consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not alter the design function or 
operation of the system involved. The CREF system will still provide 
protection to control room occupants in case of a significant 
radioactive release. The revised TS surveillance requirements 
provide an alternative test method that has been widely accepted for 
the measurement of CRE unfiltered inleakage. The proposed change 
does not introduce any new modes of plant or CREF system operation 
and does not involve physical modifications to the plant. Therefore, 
the proposed change does not create the potential for a new or 
different kind of accident from any accident previously evaluated.
    3. The (proposed) change does not involve a significant 
reduction in the margin of safety.
    The proposed change to the Fermi 2 TS surveillance requirements 
does not affect the radiological release from a design basis 
accident nor the postulated dose to the control room occupants as a 
result of the accident. The alternate surveillance test requirements 
provide an acceptable approach for the measurement of CRE inleakage. 
Safety margins and analytical conservatisms are included in the 
analyses to ensure that all postulated event scenarios are bounded. 
The proposed TS requirements continue to ensure that the 
radiological consequences at the control room are below the 
corresponding regulatory guidelines and that compliance with GDC-19 
is not affected. Therefore, the proposed changes will not result in 
a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of amendment request: October 10, 2002.
    Description of amendment request: The amendment would allow Duke 
Energy Corporation to continue using the reactor coolant system cold 
leg elbow tap flow coefficient that was approved by Nuclear Regulatory 
Commission on an interim basis for Cycle 12 at Catawba Nuclear Station, 
Unit 2. No changes in Technical Specifications are necessary for this 
Amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
operation of the facility in accordance with the proposed amendment 
would not:

[[Page 70766]]

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

First Standard
    The proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated. No 
component modification, system realignment, or change in operating 
procedure will occur which could affect the probability of any accident 
or transient. The revised cold leg elbow tap flow coefficients will not 
change the probability of actuation of any Engineered Safeguards 
Feature or other device. The actual Unit 2 RCS [reactor coolant system] 
flow rate will not change. Therefore, the consequences of previously 
analyzed accidents will not change as a result of the revised flow 
coefficients.
Second Standard
    The proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. No 
component modification or system realignment will occur which could 
create the possibility of a new event not previously considered. No 
change to any methods of plant operation will be required. The elbow 
taps are already in place, and are presently being used to monitor flow 
for Reactor Protection System purposes. They will not initiate any new 
events.
Third Standard
    The proposed amendment will not involve a significant reduction in 
a margin of safety. The removal of some of the excess flow margin, 
which was introduced by the hot leg streaming flow penalties in later 
calorimetrics, will allow additional operating margin between the 
indicated flow and the Technical Specification minimum measured flow 
limit. The proposed changes in the cold leg elbow tap flow coefficients 
will continue to be conservative with respect to the analytical model 
flow predictions, since the proposed coefficients will continue to 
contain some hot leg streaming penalties from the calorimetric 
determined coefficients used in the average.
    An increase in the RCS flow indication of approximately 1.0% will 
increase the margin to a reactor trip on low flow but will not 
adversely affect the plant response to low flow transients. Current 
UFSAR [updated final safety analysis report] chapter 15 transients that 
would be expected to cause a reactor trip on the RCS low flow trip 
setpoint are Partial Loss of Reactor Coolant Flow, Reactor Coolant Pump 
Shaft Seizure and [RCP] Reactor Coolant Pump Shaft break transients. 
Three reactor trip functions provide protection for these transients, 
RCS low flow reactor trip, RCP undervoltage reactor trip and RCP 
underfrequency reactor trip. The transient analyses of these events 
assume the reactor is tripped on the low flow reactor trip setpoint. 
This is conservative and produces a more severe transient response 
since a reactor trip on undervoltage or underfrequency would normally 
be expected to trip the reactor sooner and therefore reduce the 
severity of these transients.
    The RCS low flow reactor trip is currently set at 91% of the 
Technical Specification minimum measured flow of 390,000 gpm. The 
setpoint will not be revised as a result of this change, which means 
the transients relying on this function will behave in the same manner 
with the reactor trips occurring at essentially the same conditions as 
previously analyzed. Therefore, any small increase in the reactor trip 
margin gained by the small increase in the indicated RCS flow will not 
adversely affect the plant response during these low flow events.
    Based upon the preceding discussion, Duke Energy has concluded that 
the proposed amendment does not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

FPL Energy Seabrook, Docket No. 50-443, Seabrook Station, Unit No. 1, 
Rockingham County, New Hampshire

    Date of amendment request: October 11, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.9.4, Containment Building 
Penetrations, to permit the equipment hatch to be open during core 
alterations and/or during movement of irradiated fuel assemblies within 
containment. The appropriate TS Bases would also be changed to reflect 
the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to the Seabrook Station Technical 
Specifications (TS) 3.9.4.a, and TS 3.9.4.b do not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed. The proposed changes will modify the 
conditions of containment closure during core alterations or during 
the movement of irradiated fuel within the containment. 
Specifically, the proposed changes will permit the new containment 
outage door to stay open during core alterations or during the 
movement of irradiated fuel within the containment.
    Postulated accidents that could result in a release of 
radioactive material through the open hatch include a fuel handling 
accident that results in breaching of the fuel rod cladding, and a 
loss of residual heat removal (RHR) cooling event that leads to core 
boiling. The radiological consequences of a design basis fuel 
handling accident in containment have been evaluated assuming that 
the containment is open to the outside atmosphere. The calculated 
offsite and control room doses resulting from a fuel handling 
accident are less than the criteria specified in USNRC [U.S. Nuclear 
Regulatory Commission] NUREG-0800, ``Standard Review Plan,'' section 
15.7.4 ``Radiological Consequence of Fuel Handling Accident,'' and 
10 CFR 50, Appendix A, ``General Design Criteria for Nuclear Power 
Plants,'' GDC [General Design Criteria]-19, ``Control Room.''
    The consequence of a loss of Residual Heat Removal (RHR) is the 
potential for release of radioactivity outside of containment. 
Closing containment penetrations is the mitigating action for that 
consequence. TS 3.9.8.1 and 3.9.8.2 require that corrective actions 
be taken immediately to restore the RHR cooling as soon as possible 
if RHR loop requirements are not met (by having one RHR loop 
operable and in operation). In addition, plant operators are 
required by the TS to close all containment penetrations providing 
direct access from the containment atmosphere to the outside 
environment within 4 hours. Since the most limiting time to boil in 
this condition (during core alterations or movement of irradiated 
fuel with at least 23 feet of water above the vessel flange) is 
approximately 8.3 hours, the risk associated with the potential for 
the coolant to boil and subsequently cause a release of radioactive 
gas to the containment atmosphere (if RHR cooling was not restored) 
is minimal.
    The proposed changes to TS 3.9.4.b will add a note pertaining to 
the personnel hatch airlock within the equipment hatch. The purpose 
of this note is to provide

[[Page 70767]]

clarification that the requirements of TS 3.9.4.b do not apply to 
the subject personnel hatch airlock when the outage equipment hatch 
is installed.
    Therefore, it is concluded that these proposed [changes] to TS 
3.9.4.a and TS 3.9.4.b do not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the Seabrook Station Technical 
Specifications (TS) 3.9.4.a and 3.9.4.b do not create the 
possibility of a new or different kind of accident from any 
previously evaluated. The proposed changes will permit the equipment 
hatch to be open during core alterations and movement of irradiated 
fuel within the containment building when the containment outage 
door is installed. The installation of the door does involve a minor 
change in the present method used to isolate containment 
penetrations for containment closure. However, the present fuel 
handling analysis, which is the most limiting event, assumes that 
the containment is open to the outside atmosphere and the entire 
airborne radioactivity is instantaneously released to the outside 
environment. This analysis results in [offsite] doses that are 
within the guideline values specified in USNRC NUREG-0800, 
``Standard Review Plan,'' section 15.7.4 ``Radiological Consequence 
of Fuel Handling Accident,'' and 10 CFR 50, Appendix A, ``General 
Design Criteria for Nuclear Power Plants,'' GDC-19, ``Control 
Room.'' Therefore, the proposed changes to the TS do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in [a] margin of safety.
    The proposed changes do not involve a significant reduction in 
[a] margin of safety. The proposed change to TS 3.9.4.a will permit 
the equipment hatch to be open during core alterations and/or during 
the movement of irradiated fuel assemblies within containment when 
the containment outage door is installed and closed or capable of 
being closed. During movement of irradiated fuel assemblies within 
containment, the most severe radiological consequences result from a 
fuel handling accident. The calculated offsite and control room 
operator calculated doses are within the acceptance criteria of 
USNRC NUREG-0800, ``Standard Review Plan,'' section 15.7.4 
``Radiological Consequence of Fuel Handling Accident,'' and 10 CFR 
50, Appendix A, ``General Design Criteria for Nuclear Power 
Plants,'' GDC-19, ``Control Room.'' Therefore, the proposed changes 
to TS 3.9.4 do not result in a reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. S. Ross, Attorney, Florida Power & 
Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief (Acting): James W. Andersen.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: October 11, 2002.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.9.3, ``Refueling Operations--
Decay Time,'' to revise the time associated with the movement of 
irradiated fuel in the reactor vessel from 100 hours to 80 hours. The 
proposed change is based on reanalysis of the radiological consequences 
of a limiting design basis fuel handling accident using an 80-hour 
decay time.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to TS 3/4.9.3 does not result in a condition 
where the design, material, and construction standards that were 
applicable prior to the proposed change are altered. The probability 
of occurrence of an accident previously evaluated for Seabrook 
Station is not altered by the proposed amendment to the technical 
specifications (TSs). The accidents remain the same as currently 
analyzed in the Updated Final Safety Analysis Report (UFSAR) as a 
result of the proposed change to the decay time. The accidents 
impacted by the new decay time have been reanalyzed and the 
applicable design limits have not been exceeded. The control room 
and offsite dose consequences for fuel handling accidents have been 
reevaluated and continue to meet acceptance limits.
    Therefore based on the above discussion, it is concluded that 
the proposed revision to TS 3/4.9.3 does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change to the decay time will not create a new 
accident scenario. The analyses impacted by the revised decay time 
have been evaluated. The new analysis of the fuel handling accident 
and spent fuel pool cooling system performance demonstrates that the 
applicable acceptance criteria continues to be met. The proposed 
change will not alter the way any structure, system or component 
functions, and will not significantly alter the manner in which the 
plant is operated. There will be no significant adverse effect on 
plant operation or accident mitigation equipment.
    Since no new failure modes are created by the proposed revision 
to TS 3/4.9.3 the proposed change does not create the possibility of 
a new or different kind of accident from any that was previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The fuel handling accident in the fuel building and containment 
has been reanalyzed for a decay time of 80 hours. The spent fuel 
pool cooling performance has also been evaluated for the revised 
decay time. These analyses demonstrate that acceptance criteria are 
still met for the revised decay time as described herein. The 
results of the revised analysis show that the resulting offsite 
doses (based on a decay time period of 80 hours are comparable to 
the original doses (100-hour decay time period) and well within (< 
25%) the limiting values of 10 CFR part 100. Control room doses are 
also well within the limit of General Design Criteria 19 to 10 CFR 
part 50, Appendix A. Therefore it is concluded that the proposed 
decay time still provides sufficient margin to dose consequences 
from fuel handling and to spent fuel pool temperature limits.
    Thus, it is concluded that the proposed revision to TS 3/4.9.3 
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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief (Acting): James W. Andersen.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: October 11, 2002.
    Description of amendment request: The proposed amendment would 
eliminate the Power Range Neutron Flux High Negative Rate Reactor Trip 
function from Technical Specification (TS) 3/4.3.1, ``Reactor Trip 
System Instrumentation,'' TS 2.2.1, ``Reactor Trip System 
Instrumentation Setpoints,'' and their associated Bases. The proposed 
changes associated with elimination of the Power Range Neutron Flux 
High Negative Rate Trip function are based on the NRC-approved analysis 
provided in Westinghouse WCAP-11394-P-A, ``Methodology for the Analysis 
of the Dropped Rod Event.'' The proposed amendment would also change TS 
3/4.10.3, ``Physics Tests,'' TS

[[Page 70768]]

3/4.10.4, ``Reactor Coolant Loops,'' and TS Table 4.3-1, ``Reactor Trip 
System Instrumentation Surveillance Requirements,'' that are associated 
with certain testing activities required during STARTUP operations. The 
proposed changes to TS 3/4.10.3 are to clarify that only the reactor 
trip Low Setpoint associated with OPERABLE Power Range Neutron Flux 
instrumentation channels is required to be set at 25% of RATED THERMAL 
POWER and to reword the time interval for the Analog Channel 
Operational Test (ACOT) in surveillance requirement (SR) 4.10.3.2 from 
``within 12 hours'' to the referenced time interval specified in TS 
Table 4.3-1, Functional Unit 2.b. In correlation with the proposed 
change to extend the ACOT interval in SR 4.10.3.2, Table 4.3-1 Note 1, 
would be changed from ``if not performed in previous 31 days'' to ``if 
not performed in previous 92 days.'' The proposed change would also 
extend the ACOT interval for those Functional Units that reference TS 
Table 4.3-1 Note 1. The proposed change to TS 3/4.10.4 will delete TS 
3/4.10.4 in its entirety since the condition allowed by TS 3/4.10.4 
(i.e., natural circulation/low flow conditions) was to support the 
initial startup test program prior to commercial operation. 
Additionally, as a result of deleting TS 3/4.10.4, the footnote which 
references TS 3/4.10.4 in TS 3/4.4.1.1 is deleted as well.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes (1) to eliminate the Power Range Neutron 
Flux High Negative Rate Trip Function, (2) not lowering the Power 
Range Neutron Flux High Setpoint to the same setpoints as that of 
the Power Range Neutron Flux Low Setpoint and Intermediate Range 
reactor trip setpoint prior to conducting Physics Testing, (3) 
extension of the surveillance interval for performing the ACOT and 
TADOT [Trip Actuating Device Operational Test] for the above 
described Reactor Trip System (RTS) Functional Units, (4) 
elimination of the Special Test Exception allowing performance of 
Physics Testing under no flow conditions, and (5) the other 
editorial and Bases changes to support the aforementioned changes do 
not increase the probability or consequences of reactor core damage 
accidents resulting from events previously analyzed. The safety 
functions of other safety related systems and components, which are 
related to mitigation of these events, have not been altered. All 
other RTS and Engineered Safety Features Actuation Systems (ESFAS) 
protection functions are not affected by the proposed changes. 
Favorable plant-specific historical data as well as industry 
practice support the proposed change to extend the surveillance 
intervals for performance of the applicable ACOT or TADOT on the 
aforementioned instrumentation channels. The proposed changes do not 
adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, configuration of the facility, or 
the manner in which it is operated. The proposed changes do not 
adversely alter or prevent the ability of structures, systems, or 
components to perform their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the Seabrook Station Updated Final Safety Analysis Report 
(UFSAR).
    Removal of the negative rate trip does not change the 
probability of a rod drop accident since it does not alter the 
physical function or characteristic of the rod control system. 
Changing surveillance intervals for calibrations does not change the 
probability of an initiating event since historical performance 
demonstrates that the instrumentation settings will be within the 
assumed tolerance at the longer interval. Since the effects of the 
negative rate trip are not considered in the rod drop accident 
analysis, therefore removal of the trip will not result in an 
increase in the consequences of the rod drop accident. Changes in 
surveillance frequencies do not change the essential character of 
accident progression, thus there is no increase in the consequences.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [The proposed changes do not] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not adversely alter the design 
assumptions, conditions, or configuration of the facility or the 
manner in which the plant is operated. No credit is taken in 
Seabrook Station's safety analyses that is reliant on the Power 
Range Neutron Flux High Negative Rate Trip Function. Extending the 
aforementioned surveillance intervals and not lowering the Power 
Range Neutron Flux High Setpoint prior to physics testing do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. There are no changes to the 
source term or radiological release assumptions used in evaluating 
the radiological consequences in the Seabrook Station UFSAR. The 
proposed changes have no adverse impact on component or system 
interactions. The proposed changes will not adversely degrade the 
ability of systems, structures and components important to safety to 
perform their safety function nor change the response of any system, 
structure or component important to safety as described in the 
UFSAR. The proposed changes do not change the level of programmatic 
and procedural details of assuring operation of the facility in a 
safe manner. Since there are no changes to the design assumptions, 
conditions, configuration of the facility, or the manner in which 
the plant is operated and surveilled, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously analyzed.
    3. [The proposed changes do not] involve a significant reduction 
in a margin of safety.
    There is no adverse impact on equipment design or operation and 
there are no changes being made to the Technical Specification 
required safety limits or safety system settings that would 
adversely affect plant safety. Elimination of the Power Range 
Neutron Flux High Negative Rate Trip Function will not cause DNB 
[Departure from Nucleate Boiling] limits to be exceeded since this 
function is not credited in Seabrook Station's safety analysis. 
Eliminating the practice of lowering the Power Range Neutron Flux 
High Setpoint prior to physics testing does not involve a 
significant reduction in the margin of safety since there is 
adequate redundancy of nuclear instrumentation channels to prevent 
core damage from a positive reactivity excursion. The proposed 
changes to extend certain surveillance intervals do not reduce the 
reliability of the aforementioned trip functions to operate as 
designed nor reduce the level of programmatic or procedural controls 
associated with the aforementioned surveillance requirements. The 
negative rate trip function could, and has, caused an inadvertent 
reactor trip. Removal of this function will not reduce any perceived 
``defense-in-depth'' since the design of the core limits rod worth 
such that DNB is acceptable during a rod drop event. Additionally, 
since WCAP-11394-P-A has demonstrated that the negative rate trip is 
not considered in the safety analysis margin, removal of the NFRT is 
not considered a ``significant reduction in margin[.] `` The other 
changes are editorial/administrative in nature which support the key 
changes as mentioned above and by their nature do not involve a 
significant reduction in a margin of safety.
    Therefore, the proposed changes as described in this License 
Amendment Request do not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief (Acting): James W. Andersen.

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: September 24, 2002.

[[Page 70769]]

    Description of amendment request: The proposed change will revise 
Technical Specification (TS) Surveillance Requirement (SR) 4.0.3, to 
incorporate the approved Consolidated Line Item Improvement Program 
change associated with the TS Task Force traveler TSTF-358, revision 6, 
SR 3.0.3, ``Missed Surveillance Requirements.'' Additionally, a change 
to the Administrative Controls Section, section 6.8, is included in 
this request to include a new TS requirement for a Bases Control 
Program, consistent with the Bases Control Program presented in chapter 
5, ``Administrative Controls,'' section 5.5, ``Programs and Manuals,'' 
of the Improved Technical Specifications (ITS) for Westinghouse plants, 
NUREG 1431, revision 2. The NRC staff issued a notice of opportunity 
for comment in the Federal Register on June 14, 2001 (66 FR 32400), on 
possible amendments concerning missed surveillances, including a model 
safety evaluation and model no significant hazards consideration (NSHC) 
determination, using the consolidated line item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
September 24, 2002, with the exception of the addition of the 
incorporation of a Bases Control Program in chapter 5, ``Administrative 
Control,'' section 5.5, ``Programs and Manuals,'' of the ITS for 
Westinghouse plants, NUREG 1431, revision 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration for the changes associated with 
extending the delay period for a missed surveillance is presented 
below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function. Therefore, this change does 
not involve a significant reduction in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration for the 
proposed administrative changes, which is presented below:

    SCE&G has reviewed the proposed no significant hazards 
consideration determination (NSHCD) published in the Federal 
Register as part of the CLIIP [Consolidated Line Item Improvement]. 
SCE&G has concluded that the proposed NSHCD presented in the Federal 
Register notice is applicable to VCSNS with one exception. The 
proposed NSHCD is hereby incorporated by reference to satisfy the 
requirements of 10 CFR 50.91(a).
    The exception is that the published NSHCD does not specifically 
address the incorporation of a Bases Control Program, as one is 
already incorporated into the ITS NUREGs. Therefore, a NSHCD is 
presented for the proposed inclusion of a Bases Control Program into 
the VCSNS TS.
    In accordance with the criteria set forth in 10 CFR 50.92, SCE&G 
has evaluated these proposed Technical Specification changes and 
determined they do not represent a significant hazards 
consideration. The following is provided to support this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change provides an addition to the Administrative 
Section of TS to comply with the requirements of the Federal 
Register published notice of availability for TSTF-358, revision 6. 
This change adds a Bases Control Program to section 6.8 that is 
consistent with the Bases Control Program in NUREG 1431, revision 2.
    A bases control program will not provide for a significant 
increase in probability or consequences of an accident previously 
evaluated as there are no changes in hardware or software for the 
plant and no changes in any operating procedure. The incorporation 
of a Bases control program into the Administrative Section of TS 
will help to assure that all assumptions in the plant accident 
analysis for initial conditions, redundancy, and independence are 
maintained. This change will assure that any and all future 
revisions to the Bases section of TS will be consistently controlled 
in a manner acceptable to both the industry and the NRC.
    Therefore, this change provides for no significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change has no impact on the operation of the plant 
or changes to plant configuration. Only the manner in which VCSNS 
processes and distributes a TS Bases change will be revised and the 
controls will be similar to the majority of the industry. The NRC 
has approved the methodology used in the Bases control program, 
located in section 5.5 of the Westinghouse Standardized Technical 
Specifications, NUREG 1431, revision 2.

[[Page 70770]]

    Therefore, there is no possibility of this change creating a new 
or different kind of accident from any previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    This change provided for a standardized methodology, acceptable 
to the NRC, to assure consistent guidance for Bases changes is 
provided and the process is controlled under a TS administrative 
program. No impact to any plant hardware or safety analysis will 
occur from this proposed change. Therefore, there is no significant 
reduction in any 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: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of amendment request: November 6, 2002.
    Description of amendment request: The proposed amendment would 
revise the Browns Ferry Nuclear Plant (BFN), Units 2 and 3, Reactor 
Pressure Vessel (RPV) material surveillance program required by 10 CFR 
50, Appendix H. This program incorporates the Boiling Water Reactor 
Vessel and Internals Project (BWRVIP) Integrated Surveillance Program 
(ISP) into the BFN Units 2 and 3 licensing basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change implements a [an] integrated surveillance 
program that has been evaluated by the NRC staff as meeting the 
requirements of paragraph III.C of Appendix H to 10 CFR 50. 
Consequently, the change does not significantly increase the 
probability of any accident previously evaluated. The change 
provides the same assurance of RPV integrity. The change will not 
cause the reactor pressure vessel or interfacing systems to be 
operated outside their design or testing limits. Also, the change 
will not alter any assumptions previously made in evaluating the 
radiological consequences of accidents. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the BFN Units 2 and 3 licensing 
basis to reflect participation in the BWRVIP ISP. The proposed 
change does not involve a modification of the design of plant 
structures, systems, or components. The change will not impact the 
manner in which the plant is operated as plant operating and testing 
procedures will not be affected by the change. The change will not 
degrade the reliability of structures, systems, or components 
important to safety as equipment protection features will not be 
deleted or modified, equipment redundancy or independence will not 
be reduced, supporting system performance will not be increased, and 
increased or more severe testing of equipment will not be imposed. 
No new accident types or failure modes will be introduced as a 
result of this proposed change. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from that previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change has been evaluated as providing an 
acceptable alternative to the plant specific RPV material 
surveillance program and meets the requirements of 10 CFR 50 
Appendix H for RPV material surveillance.
    Appendix G to 10 CFR 50 describes the conditions that require 
pressure temperature (P/T) limits and provides the general bases for 
these limits. Until the results from the Integrated Surveillance 
Program become available, RG [Regulatory Guide] 1.99, revision 2 
will be used to predict the amount of neutron irradiation damage. 
The use of operating limits based on these criteria, as defined by 
applicable regulations, codes, and standards, provide reasonable 
assurance that nonductile or rapidly propagating failure will not 
occur. The P/T limits are not derived from Design Basis Accident 
(DBA) analyses. They are prescribed during normal operation to avoid 
encountering pressure, temperature, and temperature rate of change 
conditions that might cause undetected flaws to propagate and cause 
nonductile failure of the reactor coolant pressure boundary (RCPB). 
Since the P/T limits are not derived from any DBA, there are no 
acceptance limits related to the P/T limits. Rather, the P/T limits 
are acceptance limits themselves since they preclude operation in an 
unanalyzed condition.
    The proposed change will not affect any safety limits, limiting 
safety system settings, or limiting conditions of operation. The 
proposed change does not represent a change in initial conditions, 
or in a system response time, or in any other parameter affecting 
the course of an accident analysis supporting the Bases of any 
Technical Specification. Further, the proposed change does not 
involve a revision to P/T limits but rather a revision to the 
surveillance capsule withdrawal schedule for the second surveillance 
capsule. The current P/T limits were established based on adjusted 
reference temperatures for RPV beltline materials calculated in 
accordance with RG 1.99, revision 2. P/T limits will continue to be 
revised, as necessary, for changes in adjusted reference temperature 
due to changes in fluence when two or more credible surveillance 
data sets become available. When two or more credible surveillance 
data sets become available, P/T limits will be revised as prescribed 
by RG 1.99, revision 2 or other NRC approved guidance. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: October 3, 2002.
    Description of amendment request: The amendment would revise Tables 
3.3.1-1 (Reactor Trip System (RTS) Instrumentation) and 3.3.2-1 
(Engineered Safety Feature Actuation System (ESFAS) Instrumentation) of 
Limiting Conditions for Operation (LCO) 3.3.1, ``RTS Instrumentation,'' 
and 3.3.2, ``ESFAS Instrumentation,'' of the Technical Specifications. 
The proposed changes are to the steam generator (SG) water level low-
low (adverse and normal containment environment) functions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Overall protection system performance [for the proposed changes] 
will remain within the bounds of the previously performed accident 
analyses since there are no hardware changes. The design of the SG 
water level sensing equipment and the coincidence logic in the Solid 
State Protection System will be unaffected. The only physical change 
to the RTS and ESFAS instrumentation is the increased actuation 
setpoints in the NAL

[[Page 70771]]

bistable comparator cards in the 7300 Process Protection System. 
These changes have already been implemented in the field and are in 
the conservative direction, i.e., a trip actuation signal will be 
generated sooner for an event that challenges the ability of the 
steam generators to provide a heat sink. In all other regards, the 
design of the RTS and ESFAS instrumentation will be unaffected. 
These protection systems will continue to function in a manner 
consistent with the plant design basis. All design, material, and 
construction standards that were applicable prior to this amendment 
request are maintained.
    The probability and consequences of accidents previously 
evaluated in the FSAR [Callaway Final Safety Analysis Report] are 
not adversely affected because the changes to the RTS and ESFAS trip 
setpoints assure the conservative response of the affected trip 
functions, consistent with the safety analysis and licensing basis.
    The proposed changes will not affect the probability of any 
event initiators. There will be no degradation in the performance 
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident situation. 
There will be no change to normal plant operating parameters or 
accident mitigation performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the FSAR.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes, other than increased bistable 
setpoints in the adjustable bistable comparator cards that have 
already been implemented, nor are there any changes in the method by 
which any safety-related plant system performs its safety function. 
This amendment will not affect the normal method of plant operation 
or change any operating parameters. The LCO Applicability exception 
for the SG Water Level Low-Low (Normal Containment Environment) 
channels recognizes the functional design of the system that enables 
the SG Water Level Low-Low (Adverse Containment Environment) 
channels with a higher water level trip setpoint whenever the 
Containment Pressure--Environmental Allowance Modifier channels in 
the same protection sets are tripped. No performance requirements or 
response time limits will be affected.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the performance of the 7300 
Process Protection System, Nuclear Instrumentation System, or Solid 
State Protection System used in the plant protection systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not eliminate any RTS surveillance or 
alter the frequency of surveillances required by the Technical 
Specifications. The nominal Trip Setpoints specified in the 
Technical Specification Bases have already been increased in the 
conservative direction. The safety analysis limits assumed in the 
transient and accident analyses are unchanged. None of the 
acceptance criteria for any accident analysis are changed.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (F[Delta]H), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: October 3, 2002.
    Description of amendment request: The amendment would add a phrase 
to Limiting Condition for Operation (LCO) 3.1.8, ``Physics Tests 
Exceptions--Mode 2,'' of the Technical Specifications. The phrase to be 
added is that the number of required channels for certain functions in 
Table 3.3.1-1 of LCO 3.3.1, ``RTS Instrumentation,'' may be reduced 
from four to three required channels. LCO 3.1.8 applies to reactor Mode 
2 during physics tests.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance [for the proposed change] 
will remain within the bounds of the previously performed accident 
analyses since there are no permanent hardware changes. The design 
of the RTS [reactor trip system] instrumentation will be unaffected; 
only the manner in which the system is connected for short duration 
physics testing is being changed to allow the temporary bypass of 
one power range channel. The reactor protection system will continue 
to function in a manner consistent with the plant design basis since 
a sufficient number of power range channels will remain OPERABLE to 
assure the capability of protective functions, even with a 
postulated single failure. [The number of required channels for 
certain functions in Table 3.3.1-1 is only being reduced from 4 to 3 
channels.] All design, material, and construction standards that 
were applicable prior to the request are maintained.
    The proposed change will allow the temporary bypass of one power 
range neutron flux channel during the performance of low power 
physics testing in MODE 2. This results in a temporary change to the 
coincidence logic from one-out-of-three under the current TS (with a 
trip imposed on the channel used for physics testing) to two-out-of-
three under the proposed TS (the channel used for physics testing 
would be in a bypassed state). However, this two-out-of-three 
coincidence logic still supports [the] required protection and 
control system applications, while reducing plant susceptibility to 
a spurious reactor trip.
    The proposed change will not affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters or accident mitigation performance.
    The proposed change will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in 
the FSAR [Callaway Final Safety Analysis Report].
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no permanent hardware changes nor are there any 
changes in the method by which any safety-related plant system 
performs its safety function. This change will not affect the normal 
method of power operation or change any operating parameters. No 
performance requirements will be affected.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    The proposed amendment does not alter the design or performance 
of the 7300 Process Protection System, Nuclear

[[Page 70772]]

Instrumentation System (other than as discussed above), or Solid 
State Protection System used in the plant protection systems. [The 
number of the required channels is not an initiator of an accident.]
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (F[Delta]H), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    The proposed change does not eliminate any RTS surveillance or 
alter the Frequency of surveillances required by the Technical 
Specifications. The nominal RTS and Engineered Safety Features 
Actuation System (ESFAS) trip setpoints (TS Bases Tables B 3.3.1-1 
and B 3.3.2-1), RTS and ESFAS allowable values (TS Tables 3.3.1-1 
and 3.3.2-1), and the safety analysis limits assumed in the 
transient and accident analyses [(FSAR Table 15.0-4)] are unchanged. 
None of the acceptance criteria for any accident analysis is 
changed. The potential reduction in the frequency of spurious 
reactor trips would effectively increase the margin of safety or, at 
a minium, be risk-neutral.
    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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: October 3, 2002.
    Brief description of amendment request: The proposed amendment 
would revise the definition of steam generator (SG) tube inspection in 
Technical Specification 5.5.9, ``Steam Generator Tube Surveillance 
Program.'' The amendment would add a requirement for using the rotating 
pancake coil (RPC) to the H* depth in the tubesheet. The proposed 
amendment is based on the Westinghouse Topical Report WCAP-15932-P, 
``Improved Justification of Partial-Length RPC Inspection of Tube 
Joints of Model F Steam Generators of Ameren-UE Callaway Plant,'' 
revision 0, dated September 2002.
    Date of publication of individual notice in Federal Register: 
October 18, 2002 (67 FR 64422).
    Expiration date of individual notice: November 18, 2002.

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, 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 No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

    Date of application for amendment: May 6, 2002, as supplemented 
July 25, August 12, September 6, October 15, and October 31, 2002.
    Brief description of amendment: This amendment increases the 
HBRSEP2 maximum steady-state core power level from 2300 megawatts 
thermal (MWt) to 2339 MWt, an increase of approximately 1.7 percent.
    Date of issuance: November 5, 2002.
    Effective date: November 5, 2002.
    Amendment No.: 196.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56319). The July 25, August 12, September 6, October 15, and October 
31, 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 November 5, 2002.
    No significant hazards consideration comments received: No.

[[Page 70773]]

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: May 14, 2002, as supplemented by 
letter dated September 9, 2002.
    Brief description of amendment: The amendment revised Surveillance 
Requirement (SR) 4.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period was extended from the current limit of ``* * * up to 24 
hours to permit the completion of the surveillance when the allowable 
outage time limits of the ACTION requirements are less than 24 hours'' 
to ``* * * up to 24 hours or up to the limit of the specified interval, 
whichever is greater.'' In addition, the following requirement was 
added to SR 4.0.3: ``A risk evaluation shall be performed for any 
Surveillance delayed greater than 24 hours and the risk impact shall be 
managed.'' Also, a Bases Control Program was added as Technical 
Specification 6.5.14, clarifications were made to SR 4.0.1, and other 
minor changes were made to SR 4.0.3, consistent with NUREG-1432, 
revision 2, ``Standard Technical Specifications, Combustion Engineering 
Plants.''
    Date of issuance: November 1, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 246.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 23, 2002 (67 FR 
48216). The application was renoticed on October 1, 2002 (67 FR 61680).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 1, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: July 9, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification Sections 3.10.4, ``Rod Insertion Limits,'' 3.10.5, ``Rod 
Misalignment Limitations,'' and 3.10.6, ``Inoperable Rod Position 
Indicator Channels,'' to remove the cycle-specific allowances on (1) 
rod insertion limits during individual rod position indicator channel 
calibrations and (2) rod position indicator channel accuracy for 
operation at or below 50 percent power. The amendment also revises the 
control rod indicated misalignment limits.
    Date of issuance: November 7, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 7, 2002 (67 FR 
62500).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 7, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: July 8, 2002.
    Brief description of amendments: The proposed amendments would 
change Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed change adds 
two footnotes to TS Table 3.3.8.1-1, ``Loss of Power Instrumentation,'' 
Functions 1.e and 2.e, ``Degraded Voltage--Time Delay, LOCA,'' and 
makes an editorial change to the heading of TS Table 3.3.8.1-1. The 
Degraded Voltage--Time Delay, LOCA, function is currently required to 
be OPERABLE during plant configurations when the ECCS instrumentation 
that generates the Loss of Coolant Accident (LOCA) signal is not 
required to be OPERABLE. The proposed changes correct this 
inconsistency by adding two new footnotes to TS Table 3.3.8.1-i that 
modify the required OPERABILITY of the Degraded Voltage--Time Delay, 
LOCA, function.
    Date of issuance: November 12, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 155 & 141.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53986).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 12, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: October 1, 2002, as 
supplemented October 23, 2002.
    Brief description of amendments: The amendments revise the 
licensing basis as described in the Updated Final Safety Analysis 
Report to allow lifting heavier loads with the reactor building crane 
during the Unit 1 refueling outage beginning in November 2002.
    Date of issuance: November 4, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 209 & 204.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the UFSAR.
    Date of initial notice in Federal Register: October 4, 2002 (67 FR 
62270)
    The supplement dated October 23, 2002, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 4, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-254, Quad Cities Nuclear 
Power Station, Unit 1, Rock Island County, Illinois

    Date of application for amendment: May 30, 2002, as supplemented 
August 15 and October 18, 2002.
    Brief description of amendment: The amendment revises the safety 
limit minimum critical power ratio for two-loop and single-loop 
operation for Unit 1 for Cycle 18.
    Date of issuance: November 14, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-29: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45569).
    The supplements dated August 15 and October 18, 2002, provided 
additional information that clarified the application, did not change 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards

[[Page 70774]]

consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated November 14, 2002.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: November 21, 2001, as 
supplemented January 25, 2002, and August 15, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) Surveillance Requirement (SR) 4.0.3 to 
extend the delay period, before entering a Limiting Condition for 
Operation, following a missed surveillance. The delay period is 
extended from ``* * * up to 24 hours to permit completion of the 
surveillance when the allowable outage time limits of the ACTION 
requirements are less than 24 hours' to ``* * * up to 24 hours or up to 
the limit of the specified frequency, whichever is greater.'' In 
addition, the following requirement was added to SR 4.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.'' Lastly, an editorial 
change moved two sentences dealing with operability requirements from 
SR 4.0.3 to SR 4.0.1 to make the revised TS consistent with the 
Standard TS for Combustion Engineering plants.
    Date of Issuance: November 4, 2002.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 186 and 129.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58645).
    The January 25, 2002, and August 15, 2002, Supplements did not 
affect the original proposed no significant hazards determination, or 
expand the scope of the request as noticed in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 4, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit 1, Oswego County, New York

    Date of application for amendment: June 28, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications. Specifically, it revised item 9, Shutdown Cooling 
System Isolation High Area Temperature, of Table 4.6.2b, 
``Instrumentation that Initiates Primary Coolant System or Containment 
Isolation,'' changing the frequency of instrument channel test and 
instrument channel calibration from ``once during each major refueling 
outage'' to ``once per operating cycle.''
    Date of issuance: November 13, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 177.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50956).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 13, 2002.
    No significant hazards consideration comments received: No.

North Atlantic Energy Service Corporation, et al., Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: May 17, 2002, as supplemented on June 
28, July 1, August 29, and October 11, 2002.
    Description of amendment request: The amendment revises the license 
to reflect changes related to the transfer of the license for Seabrook 
Station, Unit No. 1, previously held by North Atlantic Energy Service 
Corporation (NAESCO), as the licensed operator of the facility, and 
certain co-owners of the facility, on whose behalf NAESCO is also 
acting, to FPL Energy Seabrook, LLC.
    Date of issuance: November 1, 2002.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 86.
    Facility Operating License No. NPF-86: Amendment revised the 
License.
    Date of initial notice in Federal Register: June 14, 2002 (67 FR 
40972).
    The letters dated June 28, July 1, July 24, August 29, and October 
11, 2002, provided clarifying information and did not expand the 
application beyond the scope of the notice or affect the applicability 
of the Commission's generic no significant hazards consideration 
determination pursuant to 10 CFR 2.1315.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 25, 2002.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: July 12, 2002.
    Brief description of amendment: The amendment revised the Kewaunee 
Nuclear Power Plant Technical Specification (TS) 3.1.a.3, ``Pressurizer 
Safety Valves'' to make it consistent with the Improved Standard TS to 
improve clarity. The amendment allows both pressurizer safety valves to 
be inoperable or removed while the reactor vessel head is on, provided 
the reactor coolant system (RCS) cold legs temperature is below 200 
degrees F, which is in MODE 5 configuration. During MODE 5 
configuration, the low temperature over pressure protection system is 
available and operable to protect the RCS from overpressure.
    Date of issuance: November 7, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 164.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50957).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 7, 2002.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: August 27, 2001, as 
supplemented by letter dated August 12, 2002.
    Brief description of amendments: The amendments delete Section 
6.8.4.e, ``Post-Accident Sampling,'' from the Salem Nuclear Generating 
Station, Unit Nos. 1 and 2, Technical Specifications, and License 
Condition 2.C.25, ``Post-Accident Sampling,'' for Unit 2, thereby 
eliminating the requirements to have and maintain the post-accident 
sampling program.
    Date of issuance: November 5, 2002.
    Effective date: As the date of issuance, and shall be implemented 
within 90 days.
    Amendment Nos.: 254 and 235.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55022).

[[Page 70775]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 5, 2002.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: May 8, 2002.
    Brief description of amendment: This amendment changes TS 3.7.6 to 
exclude the control room normal and emergency air handling system from 
having to include TS 3.0.4 requirements when applying the action 
requirements of Limiting Condition for Operation 3.7.6 in Modes 5 and 
6. Specifically, the change will allow operation in a manner that is 
already permitted by TS 3.7.6.
    Date of issuance: November 7, 2002.
    Effective date: November 7, 2002.
    Amendment No.: 161.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42829).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 7, 2002.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: July 25, 2002, as supplemented by letter 
dated August 30, 2002.
    Brief description of amendment: The amendment revises paragraphs in 
Section 5.0, ``Administrative Controls,'' of the Technical 
Specifications to allow the use of generic personnel titles in place of 
plant-specific personnel titles.
    Date of issuance: November 6, 2002.
    Effective date: November 6, 2002, and shall be implemented within 
30 days of the date of issuance including the approval of the Updated 
Safety Analysis Report (USAR) change request that incorporates the 
relationships between the titles in ANSI/ANS-3.1-1978 and the plant-
specific personnel titles in the USAR, as described in the licensee's 
letters of July 25 and August 30, 2002.
    Amendment No.: 149.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53993).
    The August 30, 2002, supplemental letter provided additional 
information that clarified the application, did not change the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 6, 2002.
    No significant hazards consideration comments received: No.

    Dated in Rockville, Maryland, this 18th day of November 2002.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 02-29737 Filed 11-25-02; 8:45 am]
BILLING CODE 7590-01-P