[Federal Register Volume 67, Number 209 (Tuesday, October 29, 2002)]
[Notices]
[Pages 66005-66019]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-27243]


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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, October 4, 2002, through October 17, 2002. 
The last biweekly notice was published on October 15, 2002 (67 FR 
63687).

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 Rgister 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

[[Page 66006]]

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 November 29, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714,\1\ which is 
available at the Commission's PDR, located at One White Flint North, 
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 subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest.
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
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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

[[Page 66007]]

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 Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station (PVNGS), 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of amendments request: September 6, 2002.
    Description of amendments request: The amendments would replace the 
peak linear heat safety limit, in Technical Specification (TS) 2.1.1.2, 
``Reactor Core SLs [Safety Limits],'' by a peak fuel centerline 
temperature safety limit to have a safety limit in the TSs that would 
not be exceeded during normal operation or anticipated operational 
occurrences (AOOs), in accordance with Section 50.36(c)(1)(ii)(A) of 
Title 10 of the Code of Federal Regulations (10 CFR).
    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 does not require any physical change to 
plant systems, structures, or components nor does it require any 
change in systems or plant operations. The proposed change does not 
result in any change to safety analysis methods or results. The 
change to establish peak fuel centerline temperature as the Safety 
Limit is consistent with the PVNGS Units 1, 2 and 3 licensing bases 
for ensuring that the fuel design limits are met. Operations and 
analysis will continue to be in accordance with the PVNGS Units 1, 2 
and 3 licensing bases. The peak fuel centerline temperature is the 
basis for protecting the fuel and is consistent with the safety 
analysis. [The peak linear heat rate and peak fuel centerline 
temperature safety limits are not initiators 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 change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The PVNGS Units 1, 2 and 3 Updated Final Safety Analysis Report 
(UFSAR) Chapter 15 accident analyses for AOOs where the peak linear 
heat rate may exceed the existing Safety Limit of 21 kW/ft are the 
control element assembly (CEA) Withdrawal events at Subcritical and 
Low Power conditions. The analyses for these AOOs indicate that the 
peak fuel centerline temperature is not exceeded. The existing 
safety analyses, which remain unchanged, do not affect any accident 
initiators that would create a new accident. [The peak linear heat 
rate and peak fuel centerline temperature safety limits are not 
initiators of accidents.]
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not result in any change to safety 
analysis methods or results. Therefore, by changing the Safety Limit 
from peak linear heat rate to peak fuel centerline temperature the 
margins as established in the PVNGS Units 1, 2 and 3 Technical 
Specifications and UFSAR are unchanged.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, APS [Arizona Public Service Company] 
concludes that the activities associated with the proposed 
amendment[s] 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 amendments involves no significant hazards consideration.
    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.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendment request: September 20, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3.9.5, Shutdown Cooling (SDC) and Coolant 
Circulation--Low Water Level, for Unit Nos. 1 and 2 to add two notes to 
allow operational changes in the Shutdown Cooling System to support 
operations and testing. The changes would allow the SDC pumps to be 
deenergized for less than or equal to 15 minutes when switching from 
one train to another. The second change would allow one SDC loop to be 
inoperable for up to 2 hours for surveillance testing, provided that 
the other loop was operable and in operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The system affected by this proposed amendment is the Shutdown 
Cooling (SDC) System. This system mitigates the consequences of a 
boron dilution event and removes decay heat from the Reactor Coolant 
System when the unit is in Mode 6. This proposed amendment revises 
the Technical Specification to allow the SDC pumps to be deenergized 
for less than or equal to 15 minutes to allow swapping from one 
operating train to another, and would allow one SDC loop to be 
inoperable for up to two hours for surveillance testing. Because 
this system is used for the mitigation of an accident, it is not an 
accident initiator. Therefore, the probability of an accident 
previously evaluated is not increased.
    The only design basis accident considered in this Mode is a 
boron dilution event. Consideration is also given to a loss of decay 
heat removal in this Mode as well. Both of these conditions are 
evaluated in the Updated Final Safety Analysis Report (UFSAR). The 
evaluations consider operation of the SDC system to mitigate these 
conditions. Removing this system from service for a limited amount 
of time, with other operational restrictions, limits the 
consequences to those already assumed in the UFSAR. Thus, no 
increase in offsite dose occurs under this conditions. Therefore, 
the consequences of an accident previously evaluated have not 
increased.
    Therefore, the probability or consequences of an accident 
previously evaluated have not significantly increased.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed changes do not involve a significant change in the 
operation of the plant and no new accident initiation mechanism is 
created by the proposed changes. The SDC System is not being altered 
by this amendment request. No substantial changes are made in the 
way in which the SDC System is operated. The only change made would 
allow both SDC pumps to be

[[Page 66008]]

deenergized to swap operating trains, and one SDC inoperable for 
less than two hours to allow for surveillance testing. Since the SDC 
System is an accident mitigating system only, changes in when this 
system is needed to operate cannot create a new [kind] of accident.
    Therefore, the possibility of a new or different [kind] of 
accident from any previously evaluated is not created.
    3. Would not involve a significant reduction in a margin of 
safety.
    The margin of safety provided by the SDC System is to provide 
boration control and to remove decay and sensible heat from the 
Reactor Coolant System as described in the UFSAR. Removal of system 
components from service as described above, and with limitations in 
place to prevent boron dilution and loss of decay and sensible heat 
removal, does not significantly impact the margin of safety. The SDC 
System will continue to be able to provide its safety function under 
this conditions. Operators will continue to have adequate time to 
respond to any off-normal events. Removing the system from service, 
for a limited period of time, with other operational restrictions 
limits the consequences to those already assumed in the UFSAR. 
Therefore, no reduction in [a] margin of safety has occurred because 
the event results in the UFSAR are not changed by operation in the 
proposed conditions.
    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Richard J. Laufer.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, and Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, located in Mecklenburg County, North 
Carolina and York County, South Carolina

    Date of amendment request: August 26, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for diesel fuel oil for the 
plant's onsite diesel-generator power sources. The proposed changes 
would allow the use of an optional water and sediment content test, 
would relocate the specific version of certain American Society for 
Testing and Materials (ASTM) references to licensee controlled 
documents, would add several new ASTM references, and would relocate 
the requirement for a 10-year diesel fuel oil tank inspection and 
cleaning to licensee controlled documents. The licensee stated that the 
changes are consistent with the Standard Technical Specification 
Travelers (TSTF) 374, Revision 0 and TSTF 2, Revision 1. Associated 
changes are also proposed for the TS Bases.
    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 
change 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 amendments 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

First Standard

    The proposed changes relocate the specific American Society for 
Testing and Materials (ASTM) Standard references from the 
Administrative Controls Section of Technical Specifications (TS) to 
a licensee-controlled document. Since any changes of the licensee-
controlled document will be evaluated to the requirements of 10 CFR 
50.59, ``Changes, tests, and experiments,'' no increase in the 
probability or consequences of an accident previously evaluated is 
involved. In addition, the ``clear and bright'' test used to 
establish the acceptability of new fuel oil for use prior to 
addition to the storage tanks has expanded to allow a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil. The Bases for SR 3.8.3.3 (CNS) and 3.8.3.2 (MNS) 
are revised to indicate that the API gravity is tested in accordance 
with ASTM D1298 or D287.
    Relocating the specific ASTM Standard references from the TS to 
a licensee-controlled document, allowing a water and sediment test 
to be performed to establish the acceptability of new fuel oil, and 
revising the TS Bases will not affect or degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire 
and 3.8.3.6 for Catawba are revised to remove the requirement for a 
10-year tank inspection and cleaning. This requirement will be moved 
to a licensee-controlled document. Any changes of the licensee-
controlled document will be evaluated to the requirements of 10CFR 
50.59 ``Changes, tests, and experiments,''.
    This change will not affect or degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained.
    The proposed changes do not alter or prevent the ability of 
structures, systems, or components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits. The proposed change does 
not affect the source term, containment isolation, or radiological 
release assumptions used in evaluating the radiological consequences 
of an accident previously evaluated.
    Further, the proposed changes do not increase the types and 
amounts of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Second Standard

    The proposed changes relocate the specific ASTM Standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. In addition, the ``clear and bright'' 
test used to establish the acceptability of new fuel oil for use 
prior to addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. The proposed changes revise Bases B 
3.8.3 to reference the current specific ASTM standards. The Bases 
for SRs 3.8.3.3 (CNS) and 3.8.3.2 (MNS) are revised to indicate that 
the API gravity is tested in accordance with ASTM D1298 or D287.
    In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire 
and 3.8.3.6 for Catawba are revised to remove the requirement for a 
10-year tank inspection and cleaning. This requirement will be moved 
to a licensee-controlled document. Any changes of the licensee-
controlled document will be evaluated to the requirements of 
10CFR50.59 ``Changes, tests, and experiments,''.
    The changes do not involve a physical alteration to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis or licensing basis. 
Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Third Standard

    The proposed changes relocate the specific ASTM Standard 
references from the Administrative Control Section of TS to a 
licensee-controlled document. Instituting the proposed changes will 
continue to ensure the use of the current applicable ASTM

[[Page 66009]]

Standards to evaluate the quality of both new and stored fuel oil 
designated for use in the emergency diesels. The detail associated 
with the specific ASTM Standard references is not required to be in 
the TS to provide adequate protection of the public health and 
safety, since the TS still retains the requirement for compliance 
with the applicable ASTM standard. Changes to the licensee-
controlled document are performed in accordance with the provisions 
of 10 CFR 50.59. Should it be determined that future changes involve 
a potential reduction in a margin of safety, NRC review and approval 
would be necessary prior to the implementation of the changes. This 
approach provides an effective level of regulatory control and 
provides for a more appropriate change control process.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to the addition to 
storage tanks has been expanded to allow a water and sediment 
content test to be performed to establish the acceptability of new 
fuel oil. The proposed changes revise Bases B 3.8.3 to allow 
reference to the current ASTM standard. The Bases for SR 3.8.3.3 is 
revised to indicate that the API gravity is tested in accordance 
with ASTM D1298 or D287. The level of safety of facility operation 
is unaffected by the proposed changes since there is no change in 
the intent of the TS requirements of assuring fuel oil is of the 
appropriate quality for emergency DG use.
    In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire 
and 3.8.3.6 for Catawba are revised to remove the requirement for a 
10-year tank inspection and cleaning. This requirement will be moved 
to a licensee-controlled document. Any changes of the licensee-
controlled document will be evaluated to the requirements of 
10CFR50.59 ``Changes, tests, and experiments''. The level of safety 
of the facility operation is unaffected by the proposed changes 
since there is no change in the intent of the SR to clean and 
inspect the fuel tanks.
    Therefore, the proposed changes listed above 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: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: October 15, 2001, as supplemented by 
letter dated August 27, 2002.
    Description of amendment request: The proposed amendment request 
provides additional information to support a modification to Technical 
Specification 3.4.7 and limits Reactor Coolant System activity 
permitted by the ACTION statement to 60 microcuries per gram ([mu]Ci/
gm) at all power levels. The letdown line break accident analysis in 
the Final Safety Analysis Report is also changed to reflect revised 
dose consequences. This notice supercedes the biweekly Federal Register 
notice dated November 28, 2001 (66 FR 59504), based on the original 
application dated October 15, 2001.
    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. Will the operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The proposed change to the Technical Specifications 
(TS) conservatively limits Reactor Coolant System (RCS) activity 
permitted by Action Statement 3.4.7.a to 60 [mu]Ci/gm at all reactor 
power levels. The proposed change to the Final Safety Analysis 
Report (FSAR) Section 15.6.3.1 revises the letdown line break 
accident analyses.
    The probability of a previously evaluated accident is not 
affected by this change because the pre-existing iodine spike is not 
an accident initiator and the new letdown line break accident 
analysis does not affect any plant Structure, Systems, or Component 
(SSC) but merely determines the consequences of the previously 
evaluated accident.
    This TS change is conservative in that it will reduce the 
accident consequences for events occurring at lower power levels. 
The new letdown line break accident analysis meets the original 
Safety Evaluation Report (SER) and the current Standard Review Plan 
(SRP) acceptance criteria of a small fraction of the 10 CFR [Part] 
100 limits.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will the operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The probability of a new or different accident is not 
affected by this change because the new letdown line break analysis 
does not affect any plant Structure, Systems, or Component but 
merely determines the consequences of the previously evaluated 
accident.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will the operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The TS change is more limiting in that it will reduce 
the accident consequences for events occurring at lower power 
levels.
    The new letdown line break accident analysis, assuming one 
operating charging pump, meets the original SER and current SRP 
acceptance criteria of a small fraction of the 10 CFR [Part] 100 
limits. This single pump analysis provides a suitable licensing 
basis analysis and has sufficient conservatism to accommodate two 
and three pump operating scenarios that may exist during the 
operating cycle.

    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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle 
County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear 
Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of application for amendment request: September 27, 2002.
    Description of amendment request: The proposed amendment changes 
Appendix B, ``Environmental Protection Plan (Non-Radiological),'' of 
the license by removing a parenthetical reference to a superseded 
section of 10 CFR 51.
    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 deletes a reference to a superseded section 
of 10 CFR 51, ``Environmental Protection Regulations for Domestic 
Licensing and Related Regulatory Functions,'' found in the non-
radiological Environmental Protection Plans (EPPs) for Byron 
Station, LaSalle County Station and Quad Cities Nuclear Power 
Station, Units 1 and 2. The EPP (Non-Radiological) is Appendix B to 
the Facility Operating License. The change is administrative in 
nature. No physical changes to the facilities will result from the 
proposed change. The initial conditions and methodologies used in 
accident analyses remain unchanged. The

[[Page 66010]]

proposed change does not revise or alter the design assumptions for 
systems or components used to mitigate the consequences of 
accidents. Thus, accident analyses results are not impacted by this 
proposed change.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change deletes a reference to a superseded section 
of 10 CFR 51.5. The change is administrative in nature. No physical 
or operational changes to the facilities will result from the 
proposed change.
    The proposed change does not affect the design or operation of 
any system, structure, or component (SSC) in the plant. The safety 
functions of the related SSCs are not changed in any manner, nor is 
the reliability of any SSC reduced. The change does not affect the 
manner by which the facility is operated and does not change any 
facility, structure, system, or component. No new or different type 
of equipment will be installed by this proposed change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is administrative in nature and has no 
impact on the margin of safety of any Technical Specification. There 
is no impact on safety limits or limiting safety system settings. 
The change does not affect any plant safety parameters or setpoints. 
The proposed change deletes an inaccurate reference to a section of 
10 CFR 51 that has been superseded. No physical or operational 
changes to the facility will result from the proposed changes.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of amendment request: January 16, 2002.
    Description of amendment request: The amendment would make 
administrative, editorial, and format (including repagination) changes 
to the technical specification (TS) Bases index and the Administrative 
Control section of TSs. Specifically, the amendments would relocate the 
TS Bases page listings from the TS index to a TS Bases index, and 
remove certain duplicative administrative requirements from Section 6, 
``Administrative 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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed administrative changes to the TS index and to 
Section 6 of the TSs do not result in changes being made to 
structures, systems, or components (SSCs), or to event initiators or 
precursors. Also, the proposed changes do not impact the design of 
plant systems such that previously analyzed SSCs would now be more 
likely to fail. The initiating conditions and assumptions for 
accidents described in the Updated Final Safety Analysis Report 
(UFSAR) remain as previously analyzed. Thus, the proposed changes do 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The previously analyzed SSCs are unaffected by the proposed 
changes and continue to provide assurance that they are capable of 
performing their intended design function in mitigating the effects 
of design basis accidents (DBAs). As such, the consequences of 
accidents previously evaluated in the UFSAR will not be increased 
and no additional radiological source terms are generated. 
Therefore, there will be no reduction in the capability of those 
SSCs in limiting the radiological consequences of previously 
evaluated accidents and reasonable assurance that there is no undue 
risk to the health and safety of the public will continue to be 
provided. Thus, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    Therefore, the proposed administrative changes do not 
significantly increase the probability or consequences of any 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed administrative changes do not involve physical 
changes to analyzed SSCs or changes to the modes of plant operation 
defined in the technical specification. The proposed changes do not 
involve the addition or modification of plant equipment (no new or 
different type of equipment will be installed) nor do they alter the 
design or adversely affect operation of any plant systems. No new 
accident scenarios, accident or transient initiators or precursors, 
failure mechanisms, or limiting single failures are introduced as a 
result of the proposed changes.
    The proposed administrative changes do not cause the malfunction 
of safety-related equipment assumed to be operable in accident 
analyses. No new or different mode of failure has been created and 
no new or different equipment performance requirements are imposed 
for accident mitigation. As such, the proposed changes have no 
effect on previously evaluated accidents.
    Therefore, the proposed administrative changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed administrative changes do not affect any 
previously evaluated accident. The proposed changes do not adversely 
affect the TS requirements and will continue to ensure that the 
necessary plant equipment is operable in the plant conditions where 
these systems are required to operate to mitigate a DBA as described 
in the analyses presented in the UFSAR. Thus, the proposed 
administrative, editorial, and format changes do not affect plant 
safety.
    Therefore, the proposed administrative 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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of amendment request: July 23, 2002.
    Description of amendment request: The proposed amendment would 
revise the Unit 2 reactor coolant system (RCS) pressure-temperature 
curves in Technical Specification (TS) Figures 3.4-2 and 3.4-3 and 
associated TS Bases. The revised curves will bound operation of the 
unit for the remainder of its current license duration and bound 
operation with planned license amendments to increase the power level 
at which the unit is allowed to operate.
    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

[[Page 66011]]

consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.

Probability of Occurrence of an Accident Previously Evaluated

    The proposed change will revise the RCS pressure-temperature 
curves to bound operation of the reactor for up to 32 EFPY at a 
power level of up to 3800 MW for the current fuel cycle and beyond, 
to reflect new fluence analysis methodology, to reflect the use of 
ASME [American Society of Mechanical Engineers] Code Case N-641, to 
include boltup limits, and to no longer include instrument 
uncertainty margins.
    The proposed change will not result in physical changes to 
structures, systems, or components (SSCs), or to event initiators or 
precursors. The proposed change will not affect the ability of 
personnel to control RCS [Reactor Coolant System] pressure at low 
temperatures and, thereby, ensure the integrity of the RCPB [Reactor 
Coolant Pressure Boundary]. Use of ASME Code Case N-641 will be 
approved by the NRC through approval of a Donald C. Cook Nuclear 
Plant-specific exemption to requirements in 10 CFR 50.60(a) and 10 
CFR 50, Appendix G. Therefore, the proposed revision to the RCS 
pressure-temperature curve changes will have been determined in 
accordance with NRC accepted methodologies. These methodologies 
provide adequate assurance that the reactor vessel will withstand 
the effects of normal cyclic loads due to temperature and pressure 
changes, and provide an acceptable level of protection against 
brittle failure. Additionally, the proposed changes will not impact 
the design or operation of plant systems such that previously 
analyzed SSCs will be more likely to fail. The initiating conditions 
and assumptions for accidents described in the UFSAR will remain as 
previously analyzed. Therefore, the proposed changes will not 
involve a significant increase in the probability of an accident 
previously evaluated.

Consequences of an Accident Previously Evaluated

    The proposed change does not reduce the ability of any SSC to 
limit the radiological consequences of accidents described in the 
UFSAR. The proposed change will not alter any assumptions made in 
the analysis of radiological consequences of previously evaluated 
accidents, nor does it affect the ability to mitigate these 
consequences. No new or different radiological source terms will be 
generated as a result of the proposed change. Therefore, the 
proposed changes do not involve a significant increase in the 
consequences of an accident previously evaluated.
    The format changes will improve the appearance of the affected 
pages but will not affect any requirements. In summary, the 
probability of occurrence and the consequences of an accident 
previously evaluated will not be significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not result in physical changes to SSCs. 
The proposed change will not involve the addition or modification of 
plant equipment (no new or different type of equipment will be 
installed) nor will it alter the design of any plant systems. The 
proposed change solely involves RCS pressure-temperature limits. The 
types of potential accidents associated with these limits have been 
previously identified and evaluated. No new accident scenarios, 
accident or transient initiators or precursors, failure mechanisms, 
or single failures will be introduced as a result of the proposed 
changes. No new or different modes of failure will be created. The 
format changes will improve the appearance of the affected pages but 
will not affect any requirements. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed RCS pressure-temperature curves will continue to 
provide adequate margins of protection for the RCPB. The proposed 
changes have been determined, through supporting analyses, to be in 
accordance with the methodologies and criteria set forth in the 
applicable regulations, or in accordance with technically adequate 
alternatives. Compliance with these methodologies provides adequate 
margins of safety and ensures that the RCPB will withstand the 
effects of normal cyclic loads due to temperature and pressure 
changes as well as the loads associated with postulated faulted 
events as described in the UFSAR. The format changes will improve 
the appearance of the affected pages but will not affect any 
requirements. Therefore, the proposed change will not significantly 
reduce the margin of safety.
    In summary, based upon the above evaluation, [Indiana Michigan 
Power Company] I&M has concluded that the proposed changes involve 
no significant hazards consideration under the standards set forth 
in 10 CFR 50.92(c), and, accordingly, a finding of ``no significant 
hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: September 30, 2002.
    Description of amendment request: The proposed amendment requests 
permission to change Kewaunee Nuclear Power Plant (KNPP) Facility 
Operating License DRP-43 to use an upgraded computer code for design 
basis accident containment integrity analyses. KNPP is currently 
licensed to use code for Generation of Thermal-Hydraulic Information 
for Containment (GOTHIC) version 6.0a. The proposed amendment requests 
to use GOTHIC 7.0p2 (GOTHIC 7).
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Accident analyses affected by GOTHIC have each been evaluated 
and found to show good agreement between the GOTHIC 7 analysis and 
the current analysis of record (AOR). Safety analysis results using 
GOTHIC 7 are shown to satisfy all applicable design and safety 
analysis acceptance criteria. Since GOTHIC 7 conforms to design 
bases and its results are bounded by the existing safety analyses, 
its use within limits of the bounding accident analyses will not 
cause an increase in the probability or consequences of an accident 
previously evaluated. Adherence to safety analysis acceptance 
criteria prevents use of GOTHIC 7 from creating new challenges to 
components and systems that could adversely affect their ability to 
mitigate accident consequence or diminish integrity of any fission 
product barrier.
    Thus, the requested upgrade to GOTHIC 7 with [mist diffusion 
layer] MDL modeling option will not increase probability or the 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Upgrade to GOTHIC 7 is a change in analysis methods applied to 
Kewaunee [design basis accident] DBA. Analysis methods are not 
accident initiators. GOTHIC 7 will be applied in the same manner 
currently licensed and it is consistent with current plant design 
bases and licensed accident analysis methodologies. It does not 
adversely affect any fission product barrier, nor does it alter the 
safety function of safety related systems, structures, and 
components depended upon for accident prevention or mitigation. 
Equipment important to safety will continue to function within 
design. As demonstrated by the [Numerical

[[Page 66012]]

Applications Inc.] NAI report, GOTHIC 7 yields a representation of 
expected plant response for affected design basis accidents that is 
more accurate but remains conservative. GOTHIC 7 predicted results 
for affected DBA remain bounded by the limiting analyses of record.
    Thus, the requested upgrade to GOTHIC 7 does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    Upgrade to GOTHIC 7 affects Kewaunee design basis [loss of 
coolant accident] LOCA and [main steamline break] MSLB DBA 
containment analyses. The results predicted by GOTHIC 7 for these 
DBA analyses remain within limiting design basis accidents of 
record. GOTHIC 7 accuracy and conservatism in this application has 
been verified through benchmark analyses against the current 
analyses of record, validated against recognized standard data, and 
found to be appropriate for application to Kewaunee DBA. Safety 
analysis acceptance criteria are satisfied and adherence to safety 
analysis acceptance criteria using GOTHIC 7 assures that Technical 
Specification limits will not be exceeded during normal operation.
    Thus, upgrade to GOTHIC 7 does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: September 19, 2002.
    Description of amendment request: The proposed amendment would 
delete Surveillance Requirement (SR) 4.6.B.2, ``Primary System 
Boundary--Reactor Vessel Temperature and Pressure,'' from the 
Monticello Technical Specifications (TSs) on the basis of the 
licensee's commitment to (1) relocate the current requirements to the 
Updated Safety Analysis Report (USAR) and (2) implement the Boiling 
Water Reactor Vessel and Internals Program Integrated Surveillance 
Program as approved by the Nuclear Regulatory Commission (NRC) in a 
letter dated February 1, 2002. SR 4.6.B.2 currently states: ``Test 
specimens representing the reactor vessel, base weld, and weld heat 
affected zone metal shall be installed in the reactor vessel adjacent 
to the vessel wall at the core midplane level. The material sample 
program shall conform to ASTM [American Society for Testing and 
Materials] E 185-66.'' The licensee would also make related changes to 
the TS Bases 3.6/4.6.
    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 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change relocates the requirement of the TS 
Surveillance Requirement to a Licensee controlled document and 
implements 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 [Part] 50. The proposed change of 
relocating a TS Surveillance Requirement to the Monticello USAR and 
implementing an integrated surveillance program is not considered a 
precursor or initiator of an accident previously evaluated. The 
proposed change does not impact current plant operations or the 
design function of any structure, system or component. Consequently, 
the proposed change does not significantly increase the probability 
of any accident previously evaluated.
    The proposed change provides the same assurance of Reactor 
Pressure Vessel integrity as has always been assured. The relocation 
of the TS Surveillance Requirement provides an acceptable method for 
implementing the integrated surveillance program which was evaluated 
by the NRC staff as meeting the requirements of 10 CFR [Part] 50, 
Appendix H, paragraph III.C. The relocation of the TS Surveillance 
or the implementation of an integrated surveillance program is not 
an input or consideration in any accident previously evaluated, thus 
the proposed change will not increase the probability of any such 
accident occurring. The proposed amendment does not involve any 
change to the configuration or method of operation of any plant 
equipment that is used to mitigate the consequences of an accident, 
nor does it affect any assumptions or conditions in the accident 
analysis. As a result, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, operation of the facility in accordance with the 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. No equipment 
interfaces are modified and no changes to any equipment function or 
the method of operating the equipment are being made. The proposed 
change, to relocate the TS Surveillance and implement an integrated 
surveillance program, maintains an equivalent level of RPV [reactor 
pressure vessel] material surveillance and does not introduce any 
new accident initiators. The proposed change will not change the 
design, configuration or operation of the plant.
    Therefore, operation of the facility in accordance with the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change will not involve a significant reduction 
in a margin of safety.
    The proposed amendment has been evaluated as providing an 
acceptable alternative to the plant-specific RPV material 
surveillance program that meets the requirements of the regulations 
for RPV material surveillance. The proposed change does not exceed 
or alter a design basis or safety limit. The change relocates a TS 
Surveillance Requirement and implements an integrated surveillance 
program and as such does not significantly reduce the margin of 
safety.
    Therefore, operation of the facility in accordance with 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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: September 23, 2002.
    Description of amendment request: The proposed amendments would 
change the SSES 1 and 2 Technical Specifications (TSs) by revising 
limiting condition for operation (LCO) 3.6.2.3 to add a new Condition 
B, which permits both residual heat removal (RHR) suppression pool 
cooling subsystems to be inoperable for 8 hours, rather than 
immediately initiating a unit shutdown. By making this change, the 
licensee is incorporating Technical Specifications Task Force change 
traveler number 230 into its 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

[[Page 66013]]

consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    The proposed change relaxes the Required Actions of [LCO 
3.6.2.3] by allowing 8 hours to restore one RHR suppression pool 
cooling subsystem to OPERABLE status when both subsystems have been 
determined to be inoperable. Required Actions and their associated 
Completion Times are not initiating conditions for any accident 
previously evaluated. The proposed 8 hour Completion Time provides 
some time to restore required subsystem(s) to OPERABLE status, yet 
is short enough that operating an additional 8 hours is not a 
significant risk. Consequently, this change in Required Actions does 
not significantly increase the probability of occurrence of any 
accident previously evaluated. The Required Actions in the proposed 
change have been developed to provide assurance that appropriate 
remedial actions are taken in response to the degraded condition, 
considering the operability status of the RHR Suppression Pool 
Cooling System and the capability of minimizing the risk associated 
with continued operation. As a result, the consequences of any 
accident previously evaluated are not significantly increased. 
Therefore, this 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?
    The proposed change does not involve a physical modification or 
alteration of plant equipment (no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. The Required Actions and associated Completion Times in 
the proposed change have been evaluated to ensure that no new 
accident initiators are introduced. Thus, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The relaxed Required Actions do not involve a significant 
reduction in a margin of safety. The proposed change has been 
evaluated to minimize the risk of continued operation with both RHR 
suppression pool cooling subsystems inoperable. The operability 
status of the RHR Suppression Pool Cooling System, a reasonable time 
for repair or replacement of required features, and the low 
probability of a design basis accident occurring during the repair 
period have been considered in the evaluation. Therefore, this 
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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: April 16, 2002.
    Description of amendment request: The proposed change would revise 
the Technical Specifications to delete the primary containment 
isolation valves and instrumentation associated with the permanent 
removal of the reactor vessel head spray piping.
    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 not involve a significant increase in the probability or 
consequences of an accident previously analyzed?
    Response: No.
    The proposed changes to Technical Specification Tables 3.3.2-1, 
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in 
structures, systems, or components that would affect the probability 
or consequences of any accident previously evaluated in the Hope 
Creek Updated Final Safety Analysis Report.
    The proposed changes involve eliminating piping and valves 
associated with the reactor head spray. The reactor head spray 
system was initially provided to cool down the steam dryer and 
separator during shutdown. The head spray system is not credited for 
the prevention or mitigation of any accident. Therefore, neither the 
offsite or control room radiological consequences are affected. The 
head spray piping removal and addition of a bolted flange on the 
reactor coolant pressure boundary enhances plant safety by 
eliminating a source of pipe whip and potential leakage. In 
addition, the drywell penetration will be capped and welded closed. 
This will maintain primary containment integrity and will be 
periodically tested in conjunction with the containment integrated 
leak rate test.
    Therefore, as discussed above, this modification does not 
involve a significant increase in the probability or consequences 
from any accident previously analyzed.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed?
    Response: No.
    The proposed changes to Technical Specification Tables 3.3.2-1, 
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in 
structures, systems, or components that would create a new or 
different kind of accident from any accident previously evaluated in 
the Hope Creek Updated Final Safety Analysis Report.
    The proposed change to eliminate the head spray piping and the 
addition of a bolted flange on the reactor coolant pressure boundary 
enhances plant safety by eliminating a source of pipe whip and 
potential leakage. In addition, the drywell penetration will be 
capped and welded closed. This will maintain primary containment 
integrity and will be tested in conjunction with the containment 
integrated leak rate test.
    Therefore, as discussed above, this modification does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does not involve a significant reduction in [a] margin of 
safety?
    Response: No.
    The proposed change to delete the head spray valves from Tables 
3.3.2-1, 3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 does not reduce any 
margin of safety as defined in the Technical Specifications or 
Bases. The bolted flange that will be installed on the head spray 
penetration will maintain the integrity of the reactor coolant 
pressure boundary. This flange would then be tested as part of the 
reactor pressure vessel hydrostatic test. In addition, the drywell 
penetration will be capped and welded closed. This will maintain 
primary containment integrity and will be tested as part of the 
containment integrated leak rate test.
    Accordingly, based on the above, 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Andersen, Acting.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: November 7, 2001.
    Description of amendment request: The proposed amendments would 
remove license condition 2.C.3.f from the Unit 1 operating license and 
license condition 2.C.4 from the Unit 2 operating license, and replace 
them with a commitment in Section 9.1.4.2.2.5 of the Updated Final 
Safety Analysis Report (UFSAR). Specifically, license conditions 
2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively, require NRC 
approval of the lifting devices which attach the spent fuel cask to the 
crane prior to use of the spent fuel cask crane for the purpose of 
moving

[[Page 66014]]

spent fuel casks. Subsequent to issuance of FOLs NPF-2 and NPF-8, the 
NRC issued NUREG-0612, ``Control of Heavy Loads at Nuclear Power 
Plants,'' which endorsed the use of ANSI N14.6 for the design and 
inspection of special lift devises thereby eliminating the need for 
license conditions 2.C.3.f and 2.C.4. Accordingly, SNC proposes that 
license conditions 2.C.3.f and 2.C.4 be removed from FOLs NPF-2 and 
NPF-8, respectively, and replaced with a commitment in the FNP UFSAR to 
ANSI N14.6 for the design, fabrication, testing, and quality assurance 
requirements associated with the spent fuel cask lift device.
    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 replaces license conditions 2.C.3.f and2.C.4 
to FOLs NPF-2 and NPF-8, respectively, with a commitment in the FNP 
Updated Final Safety Analysis Report (UFSAR) to the requirements of 
ANSI N14.6, as clarified by NUREG-0612, for the design, fabrication, 
testing, maintenance, and quality assurance requirements applicable 
to the spent fuel cask special lift device. The proposed change does 
not involve a physical change to or require new or different 
operability requirements for plant systems, structures, or 
components. NUREG-0612, Control of Heavy Loads at Nuclear Power 
Plants, provides methods acceptable to the NRC for assuring the safe 
handling of heavy loads. NUREG-0612 endorses the use of ANSI N14.6 
for the design, fabrication, testing, maintenance, and quality 
assurance requirements applicable to special lifting devices used to 
handle heavy loads in the proximity of safe shutdown equipment and 
irradiated spent fuel, thereby eliminating the need for license 
conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively. 
Accordingly, removal of license conditions 2.C.3.f and 2.C.4 to FOLs 
NPF-2 and NFP-8, respectively, does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change replaces license conditions 2.C.3.f and 
2.C.4 from FOLs NPF-2 and NPF-8, respectively, with a commitment in 
the FNP UFSAR to the requirements of ANSI N14.6, as clarified by 
NUREG-0612, for the design, fabrication, testing, maintenance, and 
quality assurance requirements applicable to the spent fuel cask 
special lift device. The proposed change does not involve: (1) A 
physical change to plant systems, structures or components; or (2) 
require new or different operability requirements for plant systems, 
structures, or components. SNC's commitment to the guidance provided 
in ANSI N14.6, as clarified by NUREG-0612, provides assurance that 
the spent fuel cask special lift device, in conjunction with the use 
of the single-failure proof spent fuel cask crane, will preclude the 
possibility of a cask drop 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?
    The proposed change does not involve a physical change to the 
plant or impact the operability requirements of systems, structures, 
or components considered important to safety. As stated above, the 
use of ANSI N14.6, as clarified by NUREG-0612, has been endorsed by 
the NRC in NUREG-0612. The proposed change replaces license 
conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively, 
with a commitment in the FNP UFSAR to the requirements of ANSI 
N14.6, as clarified by NUREG-0612, for the design, fabrication, 
testing, maintenance, and quality assurance requirements for the 
spent fuel cask crane special lift device. 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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear (SQN) 
Plant, Unit 1, Hamilton County, Tennessee

    Date of amendment request: March 29, 2002 (TSC 02-02) as 
supplemented by a letter dated October 10, 2002.
    Description of amendment request: The proposed amendment deletes 
several of the Unit 1 Technical Specification (TS) surveillance 
requirements (SR) contained in TS 3/4.4.5, ``Steam Generators'' (SGs), 
associated with the voltage-based SG alternative repair criteria (ARC). 
In addition the proposed changes would delete License Condition 2.C.9.d 
which references commitment letters associated with SG inspection 
activities.
    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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Tennessee Valley Authority's [TVA's] proposed TS amendment does 
not compromise limits associated with SG tube integrity. TVA's 
proposed change removes existing SG tube plugging criteria (i.e., 
ARC) from the TS and reestablishes the standard TS criteria (40 
percent through-wall criteria). This change is inherently more 
conservative.
    The proposed revision does not alter plant equipment, test 
methods or operating practices. The proposed change continues to 
provide controls for safe operation of SQN SGs within the required 
limits. The proposed change does not contribute to events or 
assumptions associated with postulated design basis accidents (i.e., 
SG tube rupture). The proposed change does not affect operator 
indicators or actions required to diagnose or mitigate a SG tube 
rupture accident. The proposed revisions continue to maintain the 
required safety functions. Accordingly, the probability of an 
accident or the consequences of an accident previously evaluated is 
not increased.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    TVA's proposed amendment removes existing repair criteria and 
incorporates the more conservative TS limit for SG tube plugging 
(i.e., plug tubes with degradation depths equal to or greater than 
40 percent through-wall). This change will not give rise to new 
failure modes. The failure of a SG tube to maintain leakage 
integrity during operation is an analyzed event in the SQN Updated 
Final Safety Analysis Report. Accordingly, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    TVA's proposed TS amendment is conservative with respect to the 
margin of safety. The margin of safety is preserved through ensuring 
structural integrity and leakage integrity of the SG tubes.
    TVA's proposed change to remove ARC from the TS does not 
compromise structural integrity or leakage integrity of SG tubes. 
The proposed change invokes the standard TS tube plugging criteria 
limit (40 percent through-wall criteria) which is inherently more 
conservative.
    The proposed change does not affect the plant conditions, 
setpoints, or safety limits that could result in precursors to 
accidents or degrade accident mitigation systems. Plant system 
safety functions are not altered by the proposed change. 
Consequently, the proposed TS revisions does not reduce 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

[[Page 66015]]

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.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: September 6, 2002 (TS 00-14).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Units 1 and 2 Technical Specification (TS) 3/
4.4.9.1, ``Pressure/Temperature [P-T] Limits, Reactor Coolant System'' 
and TS 3/4.4.12, ``Low Temperature Overpressure Protection [LTOP] 
Systems.'' The proposed amendment provides two changes to the these 
specifications as described below:
    1. The proposed change relocates the information provided in these 
TSs into a pressure temperature limit report (PTLR) format in 
accordance with U.S. Nuclear Regulatory Commission (NRC) Generic Letter 
(GL) 96-03, ``Relocation of the Pressure Temperature Limit Curves and 
Low Temperature Overpressure Protection System Limits.''
    2. The proposed change also upgrades these TSs to the standard TS 
requirements for Westinghouse plants (NUREG-1431, Revision 2). In 
addition, the Tennessee Valley Authority (TVA) proposed a change to SQN 
TS 3/4.4.9.2, ``Pressurizer,'' to relocate the requirements of this TS 
into the SQN Technical Requirements Manual (TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:
    A. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed revision does not affect plant equipment, test methods 
or operating practices. The modification to SQN TSs is consistent with 
the Standard Technical Specifications for Westinghouse Plants and 
continues to provide controls for safe operation within the required 
limits. The revised specifications provide appropriate administrative 
controls for the RCS [reactor coolant system] P-T limits and LTOP 
setpoints within the PTLR for future revisions as needed. The proposed 
changes do not contribute to events or assumptions associated with 
postulated design basis accidents (DBA). The proposed revisions 
continue to maintain the required safety functions. Accordingly, the 
probability of an accident or the consequences of an accident 
previously evaluated is not increased.
    B. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed revisions are not the result of changes to plant 
equipment, test methods, or operating practices. The proposed revision 
to the SQN RCS P-T limits, and LTOP setpoints continues to ensure that 
conservative fracture toughness margins are maintained to protect 
against reactor pressure vessel failure and overpressure conditions. 
The modified P-T limits and LTOP setpoints are based on NRC approved 
methodology in conjunction with alternative methods provided in ASME 
Code Case N-640, ``Alternative Requirement Fracture Toughness for 
Development of P-T Limit Curves for ASME [American Society of 
Mechanical Engineers] Section XI, Division 1'' and WCAP-15315, 
``Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for 
Operating PWR [pressurized water reactor] and BWR [boiling water 
reactor] Plants.''
    The proposed changes to incorporate the PTLR format is 
administrative in nature and provide controls for maintaining RCS P-T 
limits and LTOP setpoints for future revisions as needed.
    The reactor vessel P-T limits and LTOP setpoints are operational 
limits and are not considered to be contributors to the generation of 
postulated accidents. The safety functions of the associated systems 
remain unchanged and do not affect the assumptions of DBAs. The 
operational limits and setpoints continue to be governed within the 
TSs/PTLR. Accordingly, the proposed changes do not create the 
possibility of a new or different kind of accident.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    TVA's proposed TS amendment provides revised reactor pressure 
vessel P-T limits and LTOP setpoints that are within the design 
capabilities of the RCS Safety Structures, Systems and Components (SSC) 
and pressure control systems. The limits are based on conservative 
design margins that ensure that plant operation is within the design 
capacity of the reactor vessel materials. Accordingly, the function of 
the RCS to provide a fission product barrier is not compromised.
    TVA's proposed change to include revised P-T and LTOP limits does 
not result in a change to system design features. The proposed change 
does not affect plant conditions that result in precursors to accidents 
or cause degradation of accident mitigation systems. The plant system 
safety functions are not altered by the proposed change.
    The proposed changes to the P-T limits and LTOP setpoints change 
the calculations and method from that described in the current TS Bases 
to one based on ASME Code Case N-640 and WCAP-15315. The effect of this 
change is to allow plant operation with different limits while 
continuing to retain conservative margins for assuring integrity of the 
reactor vessel and the RCS. Consequently, the proposed TS revisions do 
not significantly reduce the margin of safety.
    The NRC 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 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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.

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

    Date of application for amendment: July 9, 2002.

[[Page 66016]]

    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications 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 proposed amendment also 
would revise the control rod indicated misalignment limits.
    Date of publication of individual notice in Federal Register: 
October 7, 2002 (67 FR 62500).
    Expiration date of individual notice: November 6, 2002.

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.
    Brief description of amendments: The amendments revise the 
licensing basis as described in the Updated Final Safety Analysis 
Report (UFSAR) to allow lifting heavier loads with the reactor building 
crane during the Unit 1 refueling outage beginning in November 2002.
    Date of publication of individual notice in Federal Register: 
October 4, 2002 (67 FR 62270).
    Expiration date of individual notice: November 4, 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].

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: September 11, 2001, as 
supplemented on June 27 and September 19, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications, Section 3.9, ``Refueling,'' and its corresponding bases 
to permit the continuation of core alterations during refueling 
operations with the refueling interlocks inoperable by providing 
alternate actions which will preserve the intended design function of 
the inoperable interlocks.
    Date of Issuance: October 10, 2002.
    Effective date: October 10, 2002, and shall be implemented within 
30 days of issuance.
    Amendment No.: 234.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10008). The June 27 and September 19, 2002, letters provided clarifying 
information within the scope of the original application and did not 
change the staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of 
this amendment is contained in a Safety Evaluation dated October 10, 
2002.
    No significant hazards consideration comments received: No.

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: March 13, 2002, as supplemented 
May 10, August 14, September 5, September 23, and October 4, 2002.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) for HBRSEP2 to permit selective implementation of 
alternative radiological source term and modify the TS requirement for 
movement of irradiated fuel and performing core alterations.
    Date of issuance: October 4, 2002.
    Effective date: October 4, 2002.
    Amendment No.: 195.
    Facility Operating License No. DPR-23: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21285). The May 10, August 14, September 5, September 23, and October 
4, 2002, supplements contained clarifying information only and did not 
change the initial proposed 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 October 4, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: January 31, 2002, as 
supplemented by letters dated June 12, June 25, July 22, September 16, 
and October 2, 2002.
    Brief description of amendment: This amendment increases the 
licensed power level by approximately 1.7%, from 3,833 megawatts 
thermal (MWt) to 3,898 MWt. These changes result from increased 
feedwater flow measurement accuracy to be achieved by utilizing high 
accuracy ultrasonic flow measurement instrumentation.
    Date of issuance: October 10, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 156.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15622).

[[Page 66017]]

The June 12, June 25, July 22, September 16, and October 2, 2002, 
supplemental letters provided clarifying information that did not 
change the scope of the original Federal Register notice or the 
original no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 10, 2002.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: May 30, 2001.
    Description of amendment request: The amendment revises the Cooper 
Nuclear Station's Technical Specifications (TS) 5.5.7, ``Ventilation 
Filter Testing Program (VFTP),'' reflecting a correction of an 
erroneous reference to American Society of Mechanical Engineers N510-
1980.
    Date of issuance: September 30, 2002.
    Effective date: The amendment is effective on the date of issuance, 
to be implemented within 30 days from the date of issuance.
    Amendment No.: 195.
    Facility Operating License No. DPR-46: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46480). The Commission related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 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: June 28, 2002, as supplemented 
on August 15, August 16, and October 2, 2002.
    Brief description of amendments: The amendments change the Salem 
Technical Specifications (TS) requirements for Fuel Decay Time prior to 
commencing movement of irradiated fuel. TS Limiting Condition for 
Operation 3/4.9.3, ``Decay Time,'' is revised to allow fuel movement in 
the containment to commence 100 hours after the reactor has become 
subcritical between October 15th through May 15th. Should refueling 
occur between May 16th and October 14th, the current 168 hours decay 
time limit will remain in place. These requirements are valid through 
the year 2010.
    Date of issuance: October 10, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 251 and 232.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 2002 (67 FR 
55887). The August 15, August 16, and October 2, 2002, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 10, 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: July 18, 2002.
    Brief description of amendments: These amendments change the Salem 
Technical Specifications (TSs) requirements associated with its 
containment spray nozzles. The frequency of TS Surveillance Requirement 
(SR) 4.6.2.1.d for verifying that the containment spray nozzles are 
unobstructed is changed from a fixed 10-year frequency to after 
activities that could result in nozzle blockage. In this case, PSEG 
will be required to evaluate the work performed to determine the impact 
to the containment spray system, or perform an air or smoke flow test. 
The applicable Bases pages are also revised to reflect this change.
    Date of issuance: October 10, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 252 & 233.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53989). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 10, 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: November 1, 2001, as 
supplemented on October 1, 2002.
    Brief description of amendments: The changes modify the provisions 
under which equipment may be considered operable when either its normal 
or emergency power source is inoperable. Technical Specifications (TS) 
Section 3.0.5 was deleted and additional limiting conditions for 
operation were incorporated into electrical power systems TS 3.8.1.1, 
``A.C. Sources--Operating.'' The corresponding TS Bases were modified 
accordingly. The proposed changes are consistent with the 
recommendations contained in NUREG-1431, Rev. 2, ``Standard Technical 
Specifications for Westinghouse Plants.''
    Date of issuance: October 11, 2002.
    Effective date: October 11, 2002.
    Amendment Nos.: 253 and 234.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5331). The October 1, 2002 supplement was within the scope of the 
original application and did not change the staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 11, 2002.
    No significant hazards consideration comments received: No.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: February 20, 2001.
    Brief description of amendment: The amendment eliminates the 
security plan requirements from the 10 CFR Part 50 licensed site after 
the spent nuclear fuel has been transferred to the 10 CFR Part 72 
licensed Independent Spent Fuel Storage Installation and is based in 
part on exemptions from specific requirements set forth in 10 CFR Part 
73 and 10 CFR 50.54(p).
    Date of issuance: October 10, 2002.
    Effective date: October 10, 2002, to be implemented within 30 days.
    Amendment No.: 131.
    Facility Operating License No. DPR-54: The amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15930). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 10, 2002.
    No significant hazards consideration comments received: No.

[[Page 66018]]

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

    Date of amendment request: July 25, 2002.
    Brief description of amendment: The amendment revises Surveillance 
Requirements (SRs) 3.3.1.2 and 3.3.1.3 of the technical specifications 
on the reactor trip system (RTS) instrumentation. The change to SR 
3.3.1.2 replaces the reference to the nuclear instrumentation system 
channel output by a reference to the power range channel output and 
deletes Note 1 to the SR. The change to SR 3.3.1.3 is editorial.
    Date of issuance: October 2, 2002.
    Effective date: October 2, 2002, and shall be implemented within 6 
months of the date of issuance, including the incorporation of changes 
to the Technical Specification Bases as described in the licensee's 
application dated July 25, 2002.
    Amendment No.: 148.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53992). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 2002.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Assess and Management System's 
(ADAMS) Public Electronic Reading Room on the Internet at the NRC web 
site, http://www.nrc.gov/reading-rm/adams.html. If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the NRC Public Document Room (PDR) Reference 
staff at 1-800-397-4209, 304-415-4737 or by email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By November 29, 2002, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a

[[Page 66019]]

petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested persons should consult a current copy of 
10 CFR 2.714,\2\ which is available at the Commission's PDR, located at 
One White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland, and electronically on the Internet at the NRC web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1-800-397-4209, 301-415-4737, or by e-mail to
[email protected]. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
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    \2\ ``The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitoner's interest.
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
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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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of the continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the petition for leave to intervene 
and request for hearing should also be sent to the Office of the 
General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and because of continuing disruptions in delivery of mail 
to United States Government offices, it is requested that copies be 
transmitted either by means of facsimile transmission to 301-415-3725 
or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: September 26, 2002.
    Description of amendment request: The amendments consist of a one-
time change to the Dresden Updated Final Safety Analysis Report (UFSAR) 
to state that lifting heavy loads up to and including 116 tons is 
allowed prior to and during the upcoming Dresden Unit 3 refueling 
outage number 17.
    Date of issuance: October 4, 2002.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 196 and 189.
    Facility Operating License Nos. DPR-19 and DPR-25: Amendment 
revises the UFSAR.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Joliet Herald News, dated October 1, 2002. 
The notice provided an opportunity to submit comments on the 
Commission's proposed NSHC determination. No comments have been 
received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a Safety Evaluation dated October 4, 2002.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

    Dated at Rockville, Maryland, this 18th day of October 2002.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-27243 Filed 10-28-02; 8:45 am]
BILLING CODE 7590-01-P