[Federal Register Volume 67, Number 199 (Tuesday, October 15, 2002)]
[Notices]
[Pages 63687-63706]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-25990]


-----------------------------------------------------------------------

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, September 20, 2002, through October 3, 
2002. The last biweekly notice was published on October 1, 2002 (67 FR 
61674).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By November 14, 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 63688]]

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.
---------------------------------------------------------------------------

    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2) 
regarding petitions to intervene and contentions. For the complete, 
corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April 
29, 2002.
---------------------------------------------------------------------------

    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: August 28, 2002.
    Description of amendments request: The proposed change will revise 
the expiration date of the facility operating licenses for Palo Verde 
Nuclear Generating Station Units 1, 2, and 3, to recapture low-power 
testing time.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 63689]]

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.
    Response: No.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because they do not involve a change to design configuration or 
operation of the facilities. In addition, each PVNGS [Palo Verde 
Nuclear Generating Station] unit was designed and constructed to 
ensure a 40-year service life. Design features were incorporated 
that provide for inspectability of structures, systems and 
components during the 40-year service life. Surveillance, 
inspectability and maintenance practices which have been implemented 
in accordance with the American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code and the unit Technical 
Specifications provide assurance that any degradation in plant 
safety-related equipment will be identified and corrected to provide 
continued safe operation of each unit throughout the duration of the 
applicable facility operating license.
    The largest recapture period requested by the proposed amendment 
requests is 8 months (Unit 3). This recapture period represents less 
than 1.7% of the 40-year service life of the respective unit, and is 
insignificant from an aging effects perspective. Therefore, the 
proposed amendments do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Response: No.
    The proposed amendments would revise the expiration of each 
facility operating license such that the expiration of each facility 
operating license is based upon issuance of the respective FPOL 
[full power operating license] and not upon issuance of the 
respective LPOL [low power operating license]. No physical changes 
are being made to the design features or operation of the 
facilities. Therefore, the proposed amendments do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Response: No.
    The proposed amendments would revise the expiration of each 
facility operating license such that the expiration of each facility 
operating license is based upon issuance of the respective FPOL and 
not upon issuance of the respective LPOL. No physical changes are 
being made to the design features or operation of the facilities.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary and the containment structure) to limit the 
radiological dose to the public and control room operators in the 
event of an accident. The proposed amendments to the facility 
operating licenses are administrative in nature and have no impact 
on the margin of safety and robustness provided in the design and 
construction of the facilities. In addition, the proposed amendments 
will not relax any of the criteria used to establish safety limits, 
nor will the proposed amendments relax safety system settings or 
limiting conditions of operation as defined in the Technical 
Specifications. Therefore, the proposed amendments do not result in 
a significant reduction in the margin of safety.
    Based on the above information, APS [Arizona Public Service 
Company] concludes that the proposed amendments present no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c), and, accordingly, a finding of ``no significant 
hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve 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 amendments request: June 11, 2002.
    Description of amendments request: The proposed amendment revises 
Technical Specification (TS) 3.7.11, Spent Fuel Pool Exhaust 
Ventilation System, for Units 1 and 2 to redefine the applicability of 
the TS to limit the types of fuel assemblies to which it applies. This 
proposed amendment revises TS 3.7.11 to not require the ventilation be 
operable or in operation for the movement of fuel assemblies with an 
appropriate amount of decay time. An evaluation has determined that 32 
days is adequate time to allow for sufficient radioactive decay of 
short lived isotopes resulting in no increase in offsite dose if the 
ventilation system were not operable. This change is consistent with 
changes previously approved for the Improved Standard Technical 
Specifications as described in Technical Specification Task Force--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. 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 spent fuel 
pool exhaust ventilation system (SFPEVS). This system mitigates the 
consequences of a Fuel Handling Incident (FHI) by filtering 
radioactive iodine from the air above the spent fuel pool prior to 
that air being exhausted to the environment. This limits the offsite 
dose possible from a[n] FHI. This proposed amendment revises the 
Technical Specification applicability for the SFPEVS by defining 
when the ventilation system is required to limit offsite dose due to 
a[n] FHI. 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 originating in the spent fuel 
pool is the FHI. This accident is evaluated in the Updated Final 
Safety Analysis Report. The analysis assumed credit for the 
filtration system. However, a more recent evaluation shows that 32 
days after a fuel assembly has been removed from the critical 
reactor core, adequate radioactive decay has occurred which 
compensates for the filtration of the ventilation system. Thus, no 
increase in offsite dose occurs under these 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 SFPEVS is not being altered by this amendment request. No 
changes are made in the way in which the SFPEVS is operated or in 
the way fuel is moved in the spent fuel pool. The only change made 
would allow some irradiated fuel assemblies to be moved in the spent 
fuel pool without requiring the operation of the ventilation system. 
Since no changes are being made to the operation of the SFPEVS when 
it is needed for offsite dose control and the SFPEVS is a[n] 
accident mitigating system only, changes in when this system is 
needed to operate cannot create a new type 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 SFPEVS is to limit offsite 
dose due to a[n] FHI to the limits described in the Updated Final 
Safety Analysis Report. The evaluation performed indicates that 
radioactive decay can compensate for the filtration system. Thirty-
two days after fuel occupied a critical reactor core, enough 
radioactive decay has occurred that the offsite dose from a[n] FHI 
assuming no filtration is the same as the dose determined in the 
Updated Final Safety Analysis Report. Therefore, no reduction in the 
margin of safety has occurred because the

[[Page 63690]]

offsite dose is the same as the previously approved dose limits.
    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.

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: July 17, 2002.
    Description of amendment request: The proposed amendment would 
allow the installation of up to four lead fuel assemblies (LFAs) 
Manufactured by Westinghouse Electric Company (Westinghouse) into the 
Unit 2 Cycles 15 and 16 cores. Currently, Technical Specification (TS) 
4.2.1, Fuel Assemblies, only allows fuel that is clad with either 
zircaloy or ZIRLO. The Westinghouse LFAs utilizes advance zirconium-
based material for cladding. In addition, the statements currently in 
TS 4.2.1 concerning the lead test assemblies that were allowed to be 
inserted for Unit 1 Cycles 13, 14, and 15 will be deleted.
    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.
    Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies, 
states that fuel rods are clad with either zircaloy or ZIRLO. This 
reflects the requirements of 10 CFR 50.44, 50.46, and 10 CFR Part 
50, Appendix K, which also restricts fuel rod cladding materials to 
zircaloy or ZIRLO. Calvert Cliffs Nuclear Power Plant, Inc. proposes 
to insert up to four Westinghouse fuel assemblies into Calvert 
Cliffs Unit 2 that have some fuel rods clad in zirconium alloys that 
do not meet the definition of zircaloy or ZIRLO. An exemption to the 
regulations has also been requested to allow these fuel assemblies 
to be inserted into Unit 2. The proposed change to the Calvert 
Cliffs Technical Specifications will allow the use of cladding 
materials that are not zircaloy or ZIRLO for two fuel cycles once 
the exemption is approved. To obtain approval of new cladding 
materials, 10 CFR 50.12 requires that the applicant show that the 
proposed exemption is authorized by law, is consistent with common 
defense and security, will not present an undue risk to the public 
health and safety, and is accompanied by special circumstances. The 
proposed change to the Technical Specification is effective only as 
long as the exemption is effective. In addition, the statements 
concerning the exemption for Unit 1 Cycles 13, 14, and 15 have been 
deleted, since Unit 1 Cycle 15 is completed, and therefore the 
exemption has expired. The addition of what will be an approved 
temporary exemption for Unit 2 and the deletion of an expired 
exception to Technical Specification 4.2.1 does not change the 
probability or consequences of an accident previously evaluated.
    Supporting analyses indicate that since the lead fuel assemblies 
(LFAs) will be placed in non-limiting locations, the placement 
scheme and the similarity of the advanced alloy to ZIRLO will assure 
that the behavior of the fuel rods with this alloy are bounded by 
the fuel performance and safety analyses performed for the ZIRLO 
clad fuel rods in the Unit 2 Core. Therefore, the addition of these 
advanced claddings does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with existing equipment, change equipment's function, or 
change the method of operating the equipment. The proposed change 
does not affect normal plant operations or configuration. Since the 
proposed change does not change the design, configuration, or 
operation, it could not become an accident initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind] of accident from any previously 
evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The margin of safety for the fuel cladding is to prevent the 
release of fission products. Supporting analyses indicate that since 
the LFAs will be placed in non-limiting locations, the placement 
scheme and the similarity of the advanced alloy to ZIRLO will assure 
that the behavior of the fuel rods with these alloys are bounded by 
the fuel performance and safety analyses performed for the ZIRLO 
clad fuel rods in the Unit 2 cores. Therefore, the addition of the 
advanced cladding does not involve a significant reduction in the 
margin of safety.
    The proposed change will add an approved temporary exemption to 
the Unit 2 Technical Specifications allowing the installation of up 
to four Westinghouse LFAs. The assemblies use the advanced cladding 
materials that are not specifically permitted by existing 
regulations or Calvert Cliffs' Technical Specifications. A temporary 
exemption to allow the installation of these assemblies has been 
requested. The addition of an approved temporary exemption to 
Technical Specification 4.2.1 is simply intended to allow the 
installation of the LFAs under the provisions of the temporary 
exemption. The license amendment is effective only as long as the 
exemption is effective. This amendment does not change the margin of 
safety since it only adds a reference to an approved, temporary 
exemption to the Technical Specifications.
    In addition, the words concerning the exemption for Unit 1 
Cycles 13, 14, and 15 will be deleted since Unit 1 Cycle 15 is 
completed, and therefore, the exemption has expired. This change 
does not change the margin of safety since it only deletes a 
reference to an expired exemption to the Technical Specifications.
    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: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendments request: August 6, 2002.
    Description of amendments request: The proposed amendment would 
revise Technical Specification (TS) 3.9, Refueling Operations, to 
incorporate two changes previously approved in NUREG-1432, Revision 2, 
``Combustion Engineering Improved Standard Technical Specifications'' 
dated April 2001. One change would add a note to Limiting Condition for 
Operation 3.9.3 allowing penetration flow path(s) that have direct 
access from containment atmosphere to the outside atmosphere to be 
unisolated under administrative control. The other change would replace 
the requirement in TSs 3.9.4 and 3.9.5 to ``[c]lose all containment 
penetrations providing direct access from the containment atmosphere to 
outside atmosphere'' with a set of more detailed and less restrictive 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR part 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 63691]]

consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Closing the containment penetrations is considered to be a 
mitigator of the radiological consequences of a fuel handling 
incident and a loss of SDC [Shutdown Cooling], not an initiator. 
Therefore, allowing containment penetration flow paths to be 
unisolated and the containment purge valves to be opened during 
these outage activities does not involve a significant increase in 
the probability of an accident previously evaluated.
    The consequence of a fuel handling incident is the release of 
radioactivity from Containment. The impact of the proposed change to 
the calculated offsite dose resulting from a fuel handling incident 
has been evaluated and determined to be acceptable. The fuel 
handling incident analysis assumes no containment closure. The 
amount of radioactivity that could be released as a result of the 
proposed change is bounded by the current analysis of record. 
Therefore, having containment penetration flow paths unisolated 
during core alterations and fuel handling does not involve an 
increase in the consequences of an accident previously evaluated.
    The consequences of a loss of SDC is the potential for release 
of radioactivity to the atmosphere outside Containment. Closing 
containment penetrations is a mitigator of that consequence. 
Administrative controls will be put in place to ensure that in an 
emergency containment closure can be quickly achieved. The 
containment purge system isolation valves are closed automatically 
on a containment high radiation signal and can be shut by remote 
manual operation. Therefore, the proposed changes do not involve a 
significant increase in the consequences of a loss of SDC.
    Therefore, the proposed Technical Specification changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    This requested change does not involve a significant change in 
the operation of the plant and no new accident initiation mechanism 
is created by the proposed changes. Closing containment penetrations 
is considered to be a mitigator of the radiological consequences of 
any accident in the Containment, not an initiator. The containment 
penetration flow paths are currently opened and closed during the 
course of an outage. The proposed changes allow them to remain open 
during a period when they are currently required to be closed.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The margin of safety for containment closure during core 
alteration/fuel handling is based on the amount of offsite dose 
resulting from a fuel handling incident. An offsite dose calculation 
previously approved by the Nuclear Regulatory Commission for a fuel 
handling incident assumes no containment closure, and any activity 
released from the Containment is unfiltered. The analysis will apply 
to the containment penetration flow paths that could be opened under 
administrative controls and therefore, does not involve a 
significant reduction in the margin of safety.
    The margin of safety for containment closure in the case of a 
loss of SDC is twofold: (1) The time required to close the 
Containment to prevent a radioactive release to the atmosphere 
outside Containment if SDC is lost; and (2) the ability to retain 
the pressure generated by boiling of reactor coolant as a result of 
a loss of SDC.
    Currently the Technical Specifications are vague and overly 
restrictive concerning the requirement for containment closure when 
SDC is lost. The proposed change eliminates unclear requirements and 
provides a clear way to establish containment closure that meets the 
Bases description for the Action, which is to prevent fission 
products from being released from the Containment during a loss of 
SDC incident. The containment purge isolation valves close rapidly 
on a high radiation signal or are closed by remote manual operation. 
The proposed changes do not increase the possibility of a release of 
radiation following a loss of SDC incident.
    Therefore, the ability to provide containment closure is 
maintained and the margin of safety is not significantly reduce[d] 
by this proposed activity.

    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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: August 28, 2002.
    Description of amendment request: The amendment, proposed by 
Carolina Power & Light Company to the Harris Nuclear Plant (HNP) 
Technical Specifications (TS), revises TS 6.9.1.6.2 to add analytical 
methodology references, which are used to determine core operating 
limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes incorporate additional references to 
methodologies used to evaluate core operating limits. These 
methodologies have been approved by the NRC for use in licensing 
applications. Plant structures, systems, and components will not be 
operated in a different manner as a result of these proposed changes 
and no physical modifications to equipment are involved. Adding 
these references to the Core Operating Limits Report section of 
Technical Specifications does not increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes incorporate additional references to 
methodologies used to evaluate core operating limits. These 
methodologies have been approved by the NRC for use in licensing 
applications. Plant structures, systems, and components will not be 
operated in a different manner as a result of these changes and no 
physical modifications to equipment are involved. Adding these 
references to the Core Operating Limits Report section of Technical 
Specifications does not create the possibility of a new or different 
type of accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes incorporate additional references to 
methodologies used to evaluate core operating limits. These 
methodologies have been approved by the NRC for use in licensing 
applications. Plant structures, systems, and components will not be 
operated in a different manner as a result of these changes and no 
physical modifications to equipment are involved. Adding these 
references to the Core Operating Limits Report section of Technical 
Specifications 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Allen G. Howe

[[Page 63692]]

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: August 7, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) related to safety system 
settings. Specifically, the proposed changes would revise: (1) TS 1.0 
``Definitions;'' (2) TS 2.2.1 ``Limiting Safety System Settings--
Reactor Trip System Instrumentation Setpoints;'' (3) TS 3.3.1 ``Reactor 
Trip System Instrumentation;'' (4) TS 3.3.2 ``Engineered Safety 
Features Actuation System Instrumentation;'' (5) TS 3.7.7 ``Control 
Room Emergency Ventilation System;'' (6) TS 3.8.3.1 ``Onsite Power 
Distribution--Operating.'' In addition, the appropriate TS Bases would 
be revised to conform with the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes associated with the operability 
requirements, surveillance requirements and allowed outage times 
will improve usability of the facility Technical Specifications. The 
proposed changes will clearly reflect the existing plant design for 
the Reactor Trip System (RTS), Engineered Safety Features Actuation 
System (ESFAS), Control Room, Emergency Ventilation System, and 
Electrical Power Systems Instrumentation. The proposed changes will 
also provide consistency within the individual technical 
specifications tables (e.g. Table 2.2-1, Table 3.3-1, and Table 4.3-
1). In addition, there are no hardware changes associated with the 
proposed changes. Therefore, these systems will continue to perform 
within the bounds of the previously performed accident analyses.
    The proposed changes to the operability requirements will not 
affect the instrumentation's ability to mitigate the design basis 
accidents. The proposed allowed outage times (i.e. the required 
action times) are reasonable and consistent with industry guidelines 
to ensure the affected instrumentation will be restored in a timely 
manner and provide consistency with the existing plant design. The 
design basis accidents will remain the same postulated events 
described in the Millstone Unit No. 3 Final Safety Analysis Report 
(FSAR), and the consequences of these events will not be affected. 
Therefore, the proposed changes will not increase the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. The proposed changes do not alter 
the way any structure, system, or component functions and do not 
alter the manner in which the plant is operated. The proposed 
changes do not introduce any new failure modes. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not reduce the margin of safety since 
they have no impact on any accident analysis assumption. The 
proposed changes do not decrease the scope of equipment currently 
required to be operable or subject to surveillance testing, nor do 
the proposed changes affect any instrument setpoints or equipment 
safety functions. The effectiveness of Technical Specifications will 
be maintained since the changes will not alter the operation of any 
component or system, nor will the proposed changes affect any safety 
limits or safety system settings which are credited in a facility 
accident analysis. Therefore, there is no 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Andersen, Acting.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: September 12, 2002.
    Description of amendment request: The proposed amendments would 
temporarily revise Technical Specification (TS) 3.5.2, ``Emergency Core 
Cooling System (ECCS);'' TS 3.6.6, ``Containment Spray System;'' TS 
3.7.5, ``Auxiliary Feedwater (AFW) System;'' TS 3.7.7, ``Component 
Cooling Water (CCW) System;'' TS 3.7.8, ``Nuclear Service Water System 
(NSWS);'' and TS 3.8.1, ``AC Sources--Operating'' for Catawba Nuclear 
Station, Units 1 and 2. The proposed TS changes will allow the ``A'' 
NSWS header for each unit to be taken out of service for 7 days for 
pipe replacement. This pipe replacement is scheduled to occur when 
Units 1 and 2 are at power 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:

    Catawba is currently pursuing a project to replace a portion of 
the `A' train of the nuclear service water system (NSWS) piping for 
both units. This is necessary to maintain the long-term reliability 
of the NSWS. This project represents a challenge in that it is not 
possible to isolate, drain, replace, restore and test the NSWS 
during the current TS action time frame. The purpose of this 
submittal is to request a temporary change to the existing TS for 
the systems affected during the project. This will permit an orderly 
and efficient project implementation during power operation on both 
units. The specific change is to extend the TS required action time 
from 72 hours to 168 hours.
    The following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
operation of the facility in accordance with the proposed amendment 
would not:
    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 pipe replacement project for the NSWS and proposed TS 
changes have been evaluated to assess their impact on normal 
operation of the systems affected and to ensure that the design 
basis safety functions are preserved. During the pipe replacement 
the other NSWS train will be operable and no major maintenance or 
testing will be done on the operable train. The operable train will 
be protected to help ensure it would be available if called upon.
    This pipe replacement project will enhance the long term 
structural integrity in the NSWS system. This will ensure that the 
`A' NSWS header maintains its flow margin to ensure its ability to 
comply with design basis requirements and increase the overall 
reliability for many years.
    The increased NSWS train unavailability as a result of the 
implementation of this amendment does involve a one time increase in 
the probability or consequences of an accident previously evaluated 
during the time frame the NSWS header is out of service for pipe 
replacement. Considering this small time frame for the `A' NSWS 
train outage with the increased reliability and the decrease in 
unavailability of the NSWS system in the future because of this 
project, the overall probability or consequences of an accident 
previously evaluated will decrease.
    An evaluation was performed utilizing PRA [probabilistic risk 
analysis] for

[[Page 63693]]

extending the NSWS TS time limit from 72 hours to 168 hours. The 
[CDF] core damage frequency contribution from the proposed outage 
extension is judged to be acceptable for a one-time, or rare, 
evolution. Considering the change in CDF associated with the outage 
extension in the framework of an average over a five-year period, 
the average annual contribution is considered a low-to-moderate 
increase in the CDF for consideration of permanent changes to the 
licensing basis.
    Therefore, because this is a temporary and not a permanent 
change, the time averaged risk increase is acceptable. The increase 
in the overall reliability of the NSWS along with the decreased 
unavailability in the future because of the pipe replacement project 
will result in an overall increase in the safety of both Catawba 
units. Therefore, the consequences of an accident previously 
evaluated remains unaffected and there will be minimal impact on any 
accident consequences.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed temporary TS changes do not 
affect the basic operation of the ECCS [emergency core cooling 
system], containment spray system, NSWS, AFW [auxiliary feedwater], 
CCW [component cooling water], or EDG [emergency diesel generator] 
systems. The only change is increasing the required action time 
frame from 72 hours to 168 hours (ECCS, containment spray system, 
NSWS, AFW, CCW, and EDG). During the project, contingency measures 
will be in place to provide additional assurance that the affected 
systems will be able to complete their design functions. No new 
accident causal mechanisms are created as a result of NRC approval 
of this amendment request. No changes are being made to the plant, 
which will introduce any new accident causal mechanisms.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed temporary TS amendment. During the `A' NSWS train outage, 
the affected systems will still be capable of performing their 
required functions and contingency measures will be in place to 
provide additional assurance that the affected systems will be 
maintained in a condition to be able to complete their design 
functions. No safety margins will be impacted.
    The probabilistic risk analysis conducted for this proposed 
amendment demonstrated that the CD[F] associated with the outage 
extension is judged to be acceptable for a one-time or rare 
evolution. Therefore, there is not a significant reduction in the 
margin of safety.
    Based upon the preceding discussion, Duke Energy has concluded 
that the proposed amendment for a temporary one time TS change does 
not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: September 3, 2002.
    Description of amendment request: The proposed changes to the 
Columbia Generating Station Technical Specifications (TSs) are to: (1) 
Add depleted uranium to the fuel assembly composition description in TS 
4.2.1, (2) revise TS 5.6.5.b to incorporate references to the 
analytical methods used to determine core operating limits and remove 
those that are no longer used, and (3) format the revised references as 
described in Industry/Technical Specification Task Force (TSTF) 
Traveler, TSTF-363, ``Revised Topical Report References in ITS 
[Improved Technical Specifications] 5.6.5, COLR [Core Operating Limits 
Report].''
    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.
    Assembly and core designs employing depleted uranium are 
employed in other reactors and are within the FRA-ANP fuel design 
methods and experience base. There will be no change to the 
composition of the fuel pellets (i.e., UO2) containing 
the depleted uranium except for a slight decrease in the amount of 
U235. Therefore the use of depleted uranium in the fuel 
rods does not affect the mechanical performance of the rods. Flux 
profile measurements performed on these core designs correlate with 
calculated values in a manner consistent with fuel assembly designs 
that do not include depleted uranium.
    Core operating limits are established to support Technical 
Specification 3.2, Power Distribution, requirements which ensure 
that fuel design limits are not exceeded during any conditions of 
normal operation or in the event of any Anticipated Operational 
Occurrence (AOO). The methods used to determine the core operating 
limits for each operating cycle are based on methods previously 
found acceptable by the NRC and listed in TS section 5.6.5.b. A 
change to TS section 5.6.5.b is requested to include the FRA-ANP 
methods in the list of approved methods applicable to Columbia 
Generating Station. Application of these approved methods will 
continue to ensure that acceptable operating limits are established 
to protect the fuel cladding integrity during normal operation and 
AOOs.
    The requested Technical Specification changes do not involve any 
plant modifications or operational changes that could affect system 
reliability, performance, or possibility of operator error. The 
requested changes do not affect any postulated accident precursors, 
do not affect any accident mitigation systems, and do not introduce 
any new accident initiation mechanisms.
    Therefore, these changes do not increase the probability or 
consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Assembly and core designs employing depleted uranium are within 
the capability of the NRC-approved FRA-ANP fuel design methods. 
There will be no change to the composition of the fuel pellets 
(i.e., UO2) containing the depleted uranium except for a 
slight decrease in the amount of U235. Therefore the use 
of depleted uranium in the fuel rods does not affect the mechanical 
performance of the rods.
    Changes to the methodologies listed in the TS are 
administrative. The proposed changes do not involve any new modes of 
operation, any changes to setpoints, or any plant modifications. The 
core operating limits will continue to be developed using NRC-
approved methods that account for the mixed fuel core design. The 
proposed methods do not result in any new precursors to an accident.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Assembly and core designs employing depleted uranium are within 
the capability of the NRC-approved FRA-ANP fuel design methods. 
There will be no change to the composition of the fuel pellets 
(i.e., UO2) containing the depleted uranium except for a 
slight decrease in the amount of U235. Therefore the use 
of depleted uranium in the fuel rods does not affect the mechanical 
performance of the rods.
    The core operating limits will continue to be determined using 
methodologies that have been approved by the NRC.
    On this basis, the implementation of the changes does not 
involve a significant reduction in margin of safety.


[[Page 63694]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: September 19, 2002.
    Description of amendment request: The proposed amendments would add 
a new analytical method to Technical Specifications (TS) Section 5.6.5, 
``Core Operating Limits Report (COLR).'' The proposed change supports 
the core design efforts currently in process for the upcoming Unit 2 
refueling outage scheduled to begin in January 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to LaSalle County Station, Unit 1 and Unit 2 
Technical Specifications (TS), involves reference to a new fuel 
analytical method in TS Section 5.6.5, ``Core Operating Limits 
Report (COLR).'' This code package supports the methodology 
currently being used by Framatome-ANP in the reload design and 
analysis process.
    The proposed change to TS Section 5.6.5 will add to the list of 
methods used to determine the core operating limits, the fuel 
analytical method that supports design of the LaSalle County Station 
Unit 2 Cycle 10 reload that is currently scheduled to startup on 
February 5, 2003. The addition of the approved method to TS Section 
5.6.5 has no effect on any accident initiator or precursor 
previously evaluated and does not change the manner in which the 
core is operated. The NRC approved method has been reviewed to 
ensure that the output accurately models predicted core behavior, 
has no affect on the type or amount of radiation released, and has 
no affect on predicted offsite doses in the event of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to TS Section 5.6.5 does not affect the 
performance of any LaSalle County Station structure, system, or 
component credited with mitigating any accident previously 
evaluated. The use of a new analytical method, which has been 
reviewed and approved by the NRC for the design of a core reload, 
will not affect the control parameters governing unit operation or 
the response of plant equipment to transient conditions. The 
proposed change does not introduce any new modes of system operation 
or failure mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to TS Section 5.6.5 adds the current 
analytical method for design and analysis of core reloads to the 
list of methods used to determine the core operating limits. The NRC 
has approved for use by licensees the analytical method being added. 
The proposed change does not modify the safety limits or setpoints 
at which protective actions are initiated, and does not change the 
requirements governing operation or availability of safety equipment 
assumed to operate to preserve the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, Exelon Generation Company concludes that the 
proposed amendment presents a 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, 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: August 7, 2002.
    Description of amendment request: The proposed amendments would: 
(1) Revise the surveillance frequency for air or smoke flow testing of 
containment spray nozzles, as specified in surveillance requirements 
(SRs) 4.6.2.1.d and 4.6.2.2.f, from once per 10 years to following 
maintenance which results in the potential for nozzle blockage and 
allows the use of a visual examination in lieu of an air or smoke flow 
test; (2) eliminate the SR 4.6.2.2.e.3 criteria for the river water 
flow rate through the Recirculation Spray System heat exchangers; and 
(3) make minor clarifying changes to the text in Technical 
Specification 3.3.1.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed changes to the containment spray system nozzle 
surveillance frequency, the manner in which the nozzles are verified 
to be unobstructed, and the elimination of the associated 
Recirculation Spray System (RSS) flow rate requirement does not 
introduce an initiator of any design basis accident or event. The 
proposed changes do not adversely affect accident initiators or 
precursors nor alter the configuration of the facility or the manner 
in which the plant is maintained. The river/service water system 
monitoring program ensures that the river/service water flow through 
the RSS heat exchangers will be maintained. The proposed changes to 
provide alternate wording for the P-13 function in the Reactor 
Protection System solely for clarification of the current criteria 
does not adversely affect accident initiators or precursors. Thus, 
the proposed changes do not involve a significant increase in the 
probability of an accident previously evaluated.
    The proposed changes do not alter or prevent the ability of 
structures, systems, and components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits. Introduction of foreign 
materials into the containment spray system from the exterior is 
unlikely due to the location of the spray headers, the passive 
nature of the nozzles, station foreign material controls, and the 
fact that the containment spray headers are maintained dry above the 
water level maintained in the Recirculation Water Storage Tank which 
inhibits active degradation mechanisms such as corrosion. The 
proposed amendment to eliminate the associated RSS flow rate 
requirements and the text clarification for the P-13 function do not 
introduce an initiator of any design basis accident or event. The 
proposed changes are consistent with the safety analysis assumptions 
and resultant consequences. Accident analyses potentially affected 
by the proposed change have been reviewed and none are adversely 
affected. Thus, the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated.

[[Page 63695]]

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed changes to the containment spray nozzle 
surveillance frequency, the manner in which the nozzles are verified 
to be unobstructed, the elimination of the associated RSS flow rate 
requirement, and the text clarifications for the P-13 function do 
not involve any physical alteration of the plant (i.e., no new or 
different type of equipment will be installed), subsequently no new 
or different failure modes or limiting single failures are created. 
The plant will not be operated in a different manner due to the 
proposed change. All SSCs will continue to function as currently 
designed. Thus, the proposed change does not create any new or 
different accident scenarios.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed changes to the containment spray system nozzle 
surveillance frequency, the manner in which the nozzles are verified 
to be unobstructed, the elimination of the associated RSS flow rate 
requirement, and the text clarifications for the P-13 function do 
not involve revisions to any safety limits or safety system settings 
that would adversely impact plant safety. No current setpoints are 
altered by this change. The proposed amendment does not alter the 
functional capabilities assumed in a safety analysis for any SSCs 
important to the mitigation and control of design bases accident 
conditions within the facility. The river/service water system 
monitoring program ensures that the river/service water flow through 
the RSS heat exchangers will be maintained.
    All of the applicable acceptance criteria for each of the 
analyses affected by the proposed change continue to be met. The 
conclusions of the [Updated Final Safety Analysis Report] remain 
valid. Thus, since the operating parameters and system performance 
will remain within design requirements and safety analysis, safety 
margin is maintained.

    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 Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: August 23, 2002.
    Description of amendment requests: The proposed amendments would 
revise Facility Operating Licenses (OLs) DPR-58 and DPR-74, for Unit 1 
and Unit 2, respectively, and Technical Specifications (TS) for Unit 1 
and Unit 2. The licensee proposes to delete obsolete and/or expired 
license conditions from the Unit 1 and Unit 2 OLs, and make editorial 
changes to the Unit 1 and Unit 2 OLs. Administrative changes to 
specific TS for Unit 1 and Unit 2 are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    ?1. Does the proposed change involve a significant 
increase in the probability of occurrence or consequences of an 
accident previously evaluated?
    Response: No.
    The proposed deletion of obsolete and/or expired license 
conditions from the Unit 1 and Unit 2 OLs is administrative in 
nature. The deletion of these license conditions has no impact on 
plant operations since these requirements are no longer applicable. 
The proposed TS changes, the renumbering of the Unit 2 OL pages, and 
the correction of a typographical error in the Unit 1 OL are also 
administrative in nature and do not impact CNP's current design and 
licensing basis. Since the proposed changes are administrative and 
do not impact plant operations or design, the changes do not involve 
any significant increase in the probability or the consequences of 
any accident or malfunction of equipment important to safety 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed deletion of obsolete and/or expired license 
conditions from the Unit 1 and Unit 2 OLs is administrative in 
nature. The proposed TS changes, the renumbering of the Unit 2 OL 
pages, and the correction of a typographical error in the Unit 1 OL 
are also administrative in nature. These proposed changes do not 
impact plant operations or plant equipment in any manner or involve 
a physical alteration to the plant, nor a change in the methods used 
to respond to plant transients that has not been previously 
analyzed. No new or different equipment is being installed and no 
installed equipment is being removed or operated in a different 
manner. Consequently, no new failure modes are introduced and the 
proposed administrative changes to the Unit 1 and Unit 2 OL do not 
create the possibility of a new or different kind of accident or 
malfunction of equipment important to safety from any previously 
evaluated. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed deletion of obsolete and/or expired license 
conditions from the Unit 1 and Unit 2 OLs does not affect alarm or 
trip setpoints. The proposed TS changes, the renumbering of the Unit 
2 OL pages, and the correction of a typographical error in the Unit 
1 OL are administrative in nature and do not impact the condition, 
design, or performance of any plant structure, system or component. 
Thus, the results of the accident analyses will not be affected as 
any input assumptions are protected. The format changes improve 
readability and appearance and do not alter any requirements. Thus, 
the proposed changes do not involve a significant reduction in a 
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.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: August 30, 2002.
    Description of amendment requests: The proposed amendments would 
revise the reactor trip system (RTS) and engineered safety features 
actuation system (ESFAS) Technical Specification (TS) Surveillance 
Requirements in TS 3/4.3.1 and TS 3/4.3.2, respectively, by increasing 
(1) the channel operational test surveillance intervals for analog 
channels, logic cabinets, and reactor trip breakers (RTBs), and (2) the 
completion time (CT) and bypass time (BT) for the RTBs in accordance 
with the evaluation and justifications presented in the referenced 
document, WCAP-15376, Revision 0, ``Risk-Informed Assessment of the RTS 
and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and 
Completion Times,'' dated October 2000. Additionally, the proposed 
amendments would remove Mode 2 applicability for the RTS low 
pressurizer pressure and high pressurizer water level trips and to add 
a note to TS Table 4.3-1 clarifying that channel functional testing 
requirements for the reactor trip bypass breakers are only applicable 
when they are racked in

[[Page 63696]]

and closed for bypassing an RTB. The proposed amendments would also 
make format and capitalization changes to the affected TS pages that 
improve the appearance of the TS pages, but do not affect any 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes to the STIs [surveillance test intervals] 
and RTB CT and BT reduce the potential for inadvertent reactor trips 
and spurious actuations, and therefore do not increase the 
probability of any accident previously evaluated. The proposed 
changes do not change the response of the plant to any accidents and 
have an insignificant impact on the reliability of the RTS and ESFAS 
signals. These changes satisfy the acceptance criteria specified in 
the NRC's regulatory guidance for evaluating risk-informed changes 
in RG 1.174 [``An Approach for Using Probabilistic Risk Assessment 
In Risk-Informed Decisions on Plant-Specific Changes to the 
Licensing Basis,'' dated July 1998] and RG 1.177 [``An Approach for 
Plant-Specific Risk-Informed Decisionmaking: Technical 
Specifications,'' dated August 1998]. The RTS and ESFAS will 
continue to perform their functions with high reliability as 
originally assumed in the safety analysis, and the increase in risk 
is within the acceptance criteria of existing regulatory guidance; 
therefore, there will not be a significant increase in the 
consequences of any accidents.
    The RTS and ESFAS are not accident initiators or precursors in 
the safety analysis. No new initiators are created by this activity. 
The proposed changes do not change any RTS or ESFAS setpoints, nor 
do they alter the accident mitigation function of any system, 
structure or component, design assumptions, conditions or 
configuration of the facility, or the manner in which the plant is 
operated and maintained. The proposed changes do 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 or amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new system interfaces or interactions are created. The 
proposed changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The proposed 
changes do not result in a change in the manner in which the RTS and 
ESFAS provide plant protection. The RTS and ESFAS will continue to 
have the same setpoints after the proposed changes are implemented. 
The proposed changes to STI, CT, and BT do not change any existing 
accident scenarios, do not alter assumptions made in the safety 
analysis, nor create any new or different accident scenarios.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. Redundant RTS and ESFAS trains 
are maintained, and diversity with regard to the signals that 
provide reactor trip and engineered safety features actuation is 
also maintained. All signals credited as primary or secondary, and 
all operator actions credited in the accident analyses will remain 
the same. The proposed changes will not result in plant operation in 
a configuration outside the design basis. The calculated impact on 
risk is insignificant and meets the acceptance criteria contained in 
RG 1.174 and RG 1.177.
    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 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. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: March 1, 2002.
    Description of amendment request: The proposed amendment would 
revise the containment spray nozzle inspection frequency contained in 
Technical Specification Surveillance Requirement (SR) 3.6.6.9. 
Specifically, the inspection frequency would be conducted ``[f]ollowing 
maintenance which could result in nozzle blockage,'' rather than at the 
currently specified 10-year frequency. Maintenance which could result 
in nozzle blockage is controlled by procedures which establish foreign 
material exclusion (FME) controls. The FME controls require post-
maintenance verification of system cleanliness and freedom from foreign 
materials.
    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 evaluation supports the finding that operation of 
the facility in accordance with the proposed change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change revises the surveillance frequency for 
containment spray nozzle inspections from every ten years to 
following maintenance which could result in nozzle blockage. 
Analyzed events are initiated by the failure of plant structures, 
systems or components. The containment spray system is not 
considered as an initiator of any analyzed event. The proposed 
change does not have a detrimental impact on the integrity of any 
plant structure, system or component that initiates an analyzed 
event. The proposed change will not alter the operation of, or 
otherwise increase the failure probability of any plant equipment 
that initiates an analyzed accident. As a result, the probability of 
any accident previously evaluated, is not significantly increased.
    This change does not affect the plant design. Due to the plant 
design, the spray headers are maintained dry at the level of the 
nozzles. Formation of corrosion products is unlikely due to the 
corrosion resistant materials used in spray header construction. Due 
to their location at the top of the containment, introduction of 
foreign material from sources external to the spray nozzles is 
unlikely. Since loss of foreign material control when working within 
the affected boundary is the most likely cause for obstruction, 
testing or inspection following such an occurrence would verify 
nozzle condition, and the system would be capable of performing its 
safety function. As a result, the consequences of any accident 
previously evaluated are not significantly affected.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operation. No new or different type of equipment will be installed. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety for this system is based on the capacity of 
the spray headers. Since the system is not susceptible to corrosion 
induced obstruction or obstruction

[[Page 63697]]

from sources external to the spray nozzles, and performance of 
maintenance on the system would require evaluation of the potential 
for nozzle blockage and the possible need for a test or inspection, 
the likelihood that the spray nozzles might be blocked would not be 
affected by the reduction in surveillance frequency. Therefore, the 
capacity of the system would remain unaffected. Hence, this 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: L. Raghavan.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3, Limestone County, 
Alabama

    Date of amendment request: July 31, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS). The proposed amendments 
represent a full implementation of an alternative source term (AST) for 
the Units 1, 2, and 3 operating licenses. The amendments adopt the AST 
methodology by revising the current accident source term and replacing 
it with an accident source term as prescribed in 10 CFR 50.67.
    The AST analyses were performed using the guidance provided by 
Regulatory Guide 1.183, ``Alternative Source Terms for Evaluating 
Design Basis Accidents at Nuclear Power Reactors,'' dated July 2000, 
and Standard Review Plan Section 15.0.1, ``Radiological Consequences 
Analyses Using Alternative Source Terms.'' The four limiting design 
basis accidents (DBAs) considered were the Control Rod Drop Accident, 
the Refueling Accident, the Loss of Coolant Accident, and the Main 
Steam Line Break Accident.
    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.
    The AST and those plant systems affected by implementing AST do 
not initiate DBAs. The AST does not affect the design or operation 
of the facility; rather, once the occurrence of an accident has been 
postulated, the new source term is an input to evaluate the 
consequences. The implementation of the AST has been evaluated in 
the analyses for the limiting DBAs at BFN. The equipment affected by 
the proposed change is mitigative in nature and relied upon 
following an accident. The proposed changes to the TS do revise 
certain performance requirements. However, these changes will not 
involve a revision to the parameters or conditions that could 
contribute to the initiation of a design basis accident discussed in 
Chapter 14 of the BFN Updated Final Safety Analysis Report.
    Plant specific radiological analyses have been performed and, 
based on the results of these analyses, it has been demonstrated 
that the dose consequences of the limiting events considered in the 
analyses are within the regulatory guidance provided by the NRC for 
use with the AST. This guidance is presented in 10 CFR 50.67, 
Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1. 
Therefore, the proposed amendment does not result in a significant 
increase in the consequences or a significant increase in the 
probability of any previously evaluated accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Implementation of AST does not alter any design basis accident 
initiators. These changes do not affect the design function or mode 
of operations of systems, structures, or components in the facility 
prior to a postulated accident. Since systems, structures, and 
components are operated essentially no differently after the AST 
implementation, no new failure modes are created by this proposed 
change. Therefore, the proposed license amendments will 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.
    The changes proposed are associated with a revision to the 
licensing basis for BFN. The results of accident analyses revised in 
support of the proposed change are subject to the acceptance 
criteria in 10 CFR 50.67. The analyzed events have been carefully 
selected, and the analyses supporting this submittal have been 
performed using approved methodologies. The dose consequences of 
these limiting events are within the acceptance criteria provided by 
the regulatory guidance as presented in 10 CFR 50.67, Regulatory 
Guide 1.183, and SRP 15.0.1.
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, the 
changes are considered to not result in 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 are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: August 1, 2002.
    Description of amendment request: The proposed amendment would 
revise the Browns Ferry design and licensing basis as described in 
section 14.5.2.8 of the Updated Final Safety Analysis Report (UFSAR) to 
eliminate consideration of a pressure regulator downscale failure event 
as an abnormal operational transient.
    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.
    The proposed amendment involves a change in transient analysis 
assumptions and does not change the plant or the manner in which it 
is operated. Therefore, the amendment has no affect on the 
probability of an accident. The proposed amendment is based upon 
upgrades and reliability improvements made to the main turbine 
generator electro-hydraulic control system, which render the 
analysis of a Pressure Regulator Downscale Failure event and 
consideration of the associated consequences unnecessary. Therefore, 
the proposed amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment involves a change in transient analysis 
assumptions and does not change the plant or the manner in which it 
is operated. The only event affected, the Pressure Regulator Failure 
Downscale transient, is of a type already considered. Therefore, the 
proposed amendment 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.
    The proposed amendment eliminates the consideration of the 
Pressure Regulator Downscale Failure event as an abnormal

[[Page 63698]]

operational transient based on the low likelihood of occurrence of 
such an event due to improvements in the system design of the main 
turbine electro-hydraulic control system. Other abnormal operational 
pressurization transients as described in the UFSAR will continue to 
be analyzed and ensure required margins of safety to fuel thermal 
limits are maintained. Therefore, the proposed amendment does not 
involve a significant reduction in a margin of safety. In 
conclusion, the proposed amendment does not adversely affect the 
public health and safety, and does not involve any significant 
safety hazards.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 7902.
    NRC Section Chief: Allen G. Howe.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: September 3, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC in its application dated September 
3, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: August 20, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical specifications (TSs) Table 3.3.6.1-1, Function 5.a, 
Reactor Water Cleanup (RWCU) System Isolation, Main Steam Valve Vault 
Area Temperature--High, to extend the frequency of the channel 
calibration Surveillance Requirement (SR) from 122 days to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment changes the channel calibration 
surveillance frequency from 122 days to 24 months. Under certain 
circumstances, TS SR would allow a maximum surveillance interval of 
30 months for the SR. An instrumentation calculation in

[[Page 63699]]

accordance with the guidelines of Generic Letter 91-04 has shown 
that the reliability of protective instrumentation will be preserved 
for the maximum allowable surveillance interval. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change simply extends the channel calibration 
interval of instrumentation from 122 days to 24 months and does not 
affect plant modes of operation. Hence, the change does not create 
the possibility of any new failure mechanisms. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed amendment changes the instrument channel 
calibration surveillance interval from 122 days to 24 months. An 
instrumentation calculation in accordance with the guidelines of 
Generic Letter 91-04 has shown safety margins are preserved with the 
extended surveillance interval and that the TS allowable values are 
not changed. Therefore, it is concluded that the proposed amendment 
does not involve a significant reduction in a margin of safety.
    The proposed amendment changes the instrument channel 
calibration surveillance internal from 122 days to 24 months. An 
instrumentation calculation in accordance with the guidelines of 
Generic Letter 91-04 has shown safety margins are preserved with the 
extended surveillance interval and that the TS allowable values are 
changed. Therefore, it is concluded that the Proposed amendment does 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: September 3, 2002.
    Description of amendment request: The proposed amendment would 
revise technical specifications Surveillance Requirement (SR) 3.0.3 to 
extend the delay period, before entering a Limiting Condition for 
Operation, following a missed surveillance. The delay period would be 
extended from the current limit of up to 24 hours, or up to the limit 
of the surveillance frequency interval, whichever is ``less,'' to up to 
24 hours, or up to the limit of the surveillance frequency interval, 
whichever is ``greater.'' In addition, the following requirement would 
be added to SR 3.0.3: ``A risk evaluation shall be performed for any 
Surveillance delayed greater than 24 hours and the risk impact shall be 
managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). TVA reviewed the following 
proposed NSHC determination published in the Federal Register as part 
of the Consolidated Line Item Improvement Process for Technical 
Specification Task Force item 358, and concluded in its application of 
September 3, 2002, that the proposed NSHC determination applied to 
Watts Bar.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

[[Page 63700]]

    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.

Nuclear Management Company, LLC, Docket No. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Manitowoc County, Wisconsin

    Date of amendment request: April 30, 2002.
    Brief description of amendment request: The proposed amendment 
would increase the licensed reactor core power level by 1.4 percent 
from 1518.5 MWt to 1540 MWt.
    Date of publication of individual notice in Federal Register: 
September 11, 2002 (67 FR 57630).
    Expiration date of individual notice: October 11, 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, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: August 1, 2001, and as 
supplemented by letters dated June 19 and September 9, 2002.
    Brief description of amendment: The amendment revises technical 
specification requirements that have been superseded based on the 
licensed operator training program being accredited by the Institute of 
Nuclear Power Operations, promulgation of the revised 10 CFR Part 55, 
and adoption of a systems approach to training as required by 10 CFR 
50.120.
    Date of issuance: September 24, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 154.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55009). The supplemental letters dated June 19 and September 9, 2002, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 26, 2002, as supplemented 
on August 1, 2002.
    Brief description of amendment: The amendment revised the safety 
limit minimum critical power ratio values for Cycle 19 in Section 2.1.A 
of the Technical Specifications, and made several editorial or 
administrative corrections.
    Date of Issuance: September 26, 2002.
    Effective date: September 26, 2002, and shall be implemented before 
Cycle 19 startup.
    Amendment No.: 233.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50949). The August 1, 2002, letter 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 September 26, 2002.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: August 1, 2001 as supplemented 
by letters dated June 19, July 19, and September 9, 2002.
    Brief description of amendment: The amendment revised, clarified, 
and deleted, as appropriate, requirements regarding Facility Staff 
Qualifications and licensed operator and non-licensed personnel 
training programs. The changes revised requirements that have been 
superseded based on licensed operator training programs being 
accredited by the Institute for Nuclear Power Operations, promulgation 
of the revised 10 CFR Part 55, ``Operator's Licenses,'' which became 
effective on May 26, 1987, and adoption of a systems

[[Page 63701]]

approach to training as required by 10 CFR 50.120, ``Training and 
qualification of nuclear power plant personnel.''
    Date of issuance: September 23, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 241.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55009). Exelon's June 19, July 19, and September 9, 2002, letters 
provided clarifying information within the scope of the original 
application and did not change the NRC staff's proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 23, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 23, 2002, as supplemented 
July 16, 2002.
    Brief description of amendment: The amendment eliminates the 
requirement to perform response time testing for two reactor protection 
system functions and two primary containment isolation functions.
    Date of issuance: October 2, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 151.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42818). The July 16, 2002, supplemental letter provided additional 
clarifying information that was within the scope of the original 
application and did not change the Nuclear Regulatory Commission 
staff's initial proposed no significant hazards consideration 
determination.
    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.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423 Millstone Power Station, Unit Nos. 1, 2, and 3 New London County, 
Connecticut

    Date of application for amendment: November 8, 2001, as 
supplemented August 14, 2002.
    Brief description of amendment: The amendments incorporate 
administrative and editorial changes into the Millstone Unit No. 1 
Permanently Defueled Technical Specifications (PDTS) and the Millstone 
Unit Nos. 2 and 3 Technical Specifications (TSs).
    Date of issuance: September 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 111, 270 and 212.
    Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: This 
amendment revises the Unit No. 1 PDTS and the Units 2 and 3 TSs.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64290). The August 14, 2002, letter provided clarifying information 
that did not change the scope of the proposed action or the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 17, 2002.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: October 7, 2001, as 
supplemented by letter dated August 7, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications 5.6.5.a by adding a few parameter limits 
currently included in the Core Operating Limits Report.
    Date of issuance: October 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 202 and 195.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2002 (67 FR 
54680). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 1, 2002.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: October 7, 2001, as 
supplemented by letter dated August 7, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications 5.6.5.a by adding a few parameter limits 
currently included in the Core Operating Limits Report.
    Date of issuance: October 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 208 & 189.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2002 (67 FR 
54680). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 1, 2002.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: January 10, 2002.
    Brief description of amendment: The amendment revised the technical 
specifications to extend the surveillance test interval of certain 
instrument channels from the current 18 months to 24 months.
    Date of issuance: September 20, 2002.
    Effective date: September 20, 2002, to be implemented within 30 
days from the date of issuance.
    Amendment No.: 179.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50951). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 20, 2002.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: September 24, 2001, as supplemented by 
letters dated April 22 and July 29, 2002.
    Brief description of amendment: The amendment extends the allowed 
outage time for a Division I or Division II

[[Page 63702]]

Emergency Diesel Generator (EDG) from 72 hours to 14 days. The changes 
are intended to provide flexibility in scheduling EDG maintenance 
activities, reduce refueling outage duration, and improve EDG 
availability during plant shutdowns.
    Date of issuance: September 25, 2002.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 125.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64292). The April 22 and July 29, 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 September 25, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: January 31, 2002, as supplemented by 
letter dated September 9, 2002.
    Brief description of amendment: The amendment changed 
administrative Technical Specification 5.5.16 regarding the Containment 
Integrated Leak Rate Testing (ILRT) to allow a one-time extension of 
the interval (to 15 years) for performance of the next ILRT.
    Date of issuance: September 24, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 219.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7417). The September 09, 2002, supplemental letter provided 
clarifying information that did not change the scope of the original 
Federal Register notice or the original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 2002.
    No significant hazards consideration comments received: No.

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. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: August 1, 2001, as supplemented 
by letters dated June 19 and September 9, 2002.
    Brief description of amendments: The amendments would revise 
requirements that have been superseded based on licensed operator 
training programs being accredited by the Institute for Nuclear Power 
Operations, promulgation of the revised 10 CFR Part 55, ``Operators'' 
Licenses,'' which became effective on May 26, 1987, and adoption of a 
systems approach to training as required by 10 CFR 50.120, ``Training 
and qualification of nuclear power plant personnel.''
    Date of issuance: September 24, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 130 and 125.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55018). The supplements dated June 19 and September 09, 2002, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 24, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 15, 2002, as supplemented 
July 8, 2002.
    Brief description of amendments: The amendments change Technical 
Specification surveillance requirements and allowable values for 
reactor vessel steam dome pressure--high instrumentation to reflect 
replacement of pressure switches with analog units.
    Date of issuance: October 2, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days for Unit 3 and prior to startup from the next refueling 
outage for Unit 2.
    Amendment Nos.: 195 & 188.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36930). The supplement dated July 8, 2002, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 2, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: May 14, 2002, as supplemented 
by letter dated September 5, 2002.
    Brief description of amendments: These amendments revised technical 
specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the 
delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance. A TS Bases Control Program is added to 
the TSs. Additionally, two administrative changes affecting TS Section 
6.2.2, ``Unit Staff,'' and Section 6.5.1.2, ``Composition,'' were 
incorporated.
    Date of issuance: October 2, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 162 and 124.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 27, 2002 (67 FR 
55041). The supplement dated September 5, 2002, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register on August 27, 2002 (67 FR 55041).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 2, 2002.
    No significant hazards consideration comments received: No.

[[Page 63703]]

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
277, Peach Bottom Atomic Power Station, Unit 2, York County, 
Pennsylvania

    Date of application for amendment: June 10, 2002, as supplemented 
August 2, 2002.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TSs) for the safety limit for the minimum 
critical power ratio from its current value of 1.09 to 1.07 for two 
recirculation-loop operation, and from 1.10 to 1.09 for single 
recirculation-loop operation.
    Date of issuance: September 23, 2002.
    Effective date: As of the date of issuance, to be implemented prior 
to startup for cycle 15 operations, scheduled for September 2002.
    Amendment No.: 246.
    Facility Operating License No. DPR-44: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50953). The August 2, 2002, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 23, 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: January 21, 2002.
    Brief description of amendment: The amendment modifies the 
Technical Specification Surveillance Requirement 3.7.3.1 to improve 
consistency with Cooper Nuclear Station (CNS) Amendment No. 185, 
approved on March 13, 2001, and eliminate unnecessary restrictions 
regarding how the reactor equipment cooling system surge tank level is 
monitored.
    Date of issuance: September 18, 2002.
    Effective date: September 18, 2002.
    Amendment No.: 194
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15624). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 18, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: June 24, 2002.
    Brief description of amendment: The amendment revised Surveillance 
Requirement 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period was extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``* * *up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement was added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: October 2, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 107.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53987). The staff's related evaluation of the amendment is contained in 
a Safety Evaluation dated October 2, 2002.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: December 19, 2001, as 
supplemented April 19, 2002.
    Brief description of amendment: The amendment extends the time for 
completing required action A.1 of TS 3.8.4, ``Electrical Power 
Systems--DC Sources--Operating,'' for restoring the 125 volt direct 
current (VDC) electrical power subsystem to operable status. The 
change, in effect, provides for replacement of 125 VDC batteries 1D1 
and 1D2 while the plant is at power. The time is extended on a one-time 
basis, and for each battery division separately, from 8 hours to 10 
days. The one-time change also requires that required features be 
declared inoperable when the associated 125 VDC source is inoperable 
and the redundant required features are also inoperable for at least 4 
hours.
    Date of issuance: October 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 247.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5329). The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 1, 2002.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: February 8, 2002, as 
supplemented June 21, 2002.
    Brief description of amendment: Amendment changes Technical 
Specification 5.0 to be consistent with Technical Specifications Task 
Force Change No. 258, Revision 4, ``Changes to Section 5.0, 
Administrative Controls.''
    Date of issuance: October 2, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 248.
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15625). The supplemental letter contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 2, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 7, 2002, as supplemented 
August 20 and 29, 2002.
    Brief description of amendment: The amendment would revise the 
Kewaunee Nuclear Power Plant Technical Specification (TS) Sections for 
administrative changes:
    (1) Section 1-``Definitions,'' (2) Section 2-``Safety Limits and 
Limiting Safety System Settings,'' (3) Section 5-``Design Features,'' 
and (4) Section 6-``Administrative Controls.''

[[Page 63704]]

    The administrative changes include capitalizing defined words, 
formatting section titles, renumbering pages and correcting 
miscellaneous grammar and punctuation errors.
    Date of issuance: September 19, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 162.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 23, 2002 (67 FR 
48220). The supplemental letter contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 19, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 7, 2002, as supplemented 
July 19 and September 11, 2002.
    Brief description of amendment: The amendment revised the Kewaunee 
Nuclear Power Plant technical specification (TS) requirements for 
meeting surveillances in TS 4.0.a, TS requirements for missed 
surveillances in TS 4.0.c, and TS requirements for a Bases control 
program consistent with TS Bases Control Program described in Section 
5.5 of NUREG-1431, Standard TS for Westinghouse Plants, Revision 2.
    Date of issuance: September 24, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 163.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42829). The supplements dated July 19 and September 11, 2002, provided 
clarifying information that did not change the scope of the May 7, 
2002, application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 21, 2001, as 
supplemented April 26, 2002.
    Brief description of amendment: The amendment revises Technical 
Specifiction (TS) Sections 3.7/4.7, ``Containment Systems,'' to (1) 
clarify existing requirements, (2) make editorial changes, (3) revise 
limiting conditions for operation (LCOs) and surveillance requirements, 
and (4) add certain LCOs.
    Date of issuance: September 23, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 130.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34490). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 23, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 17, 2001, as 
supplemented June 25, 2002.
    Brief description of amendment: The amendment revises the 
multiplier values for the single-loop operation average planar linear 
heat generation rate to account for the use of General Electric (GE)14 
fuel.
    Date of issuance: October 2, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 131.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57122). 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.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: September 13, 2001, as 
supplemented by letter dated February 27, 2002.
    Brief description of amendments: The amendment revises Technical 
Specification (TS) 3.7.16, ``Spent Fuel Pool Boron Concentration''; TS 
3.7.17, ``Spent Fuel Assembly Storage--Region 1/Region 2''; and TS 4.3, 
``Fuel Storage,'' for Diablo Canyon Nuclear Power Plant Units 1 and 2, 
to allow the use of credit for soluble boron in the spent fuel pool 
criticality analysis.
    Date of issuance: September 25, 2002.
    Effective date: September 25, 2002, and shall be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-154; Unit 2-154.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55020). The February 27, 2002, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 25, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M. 
Farley Nuclear Plant, Unit 1, Houston County, Alabama

    Date of amendment request: March 4, 2002, as supplemented by letter 
dated July 11, 2002.
    Brief Description of amendment: The proposed amendment revises 
Technical Specifications (TS) 5.5.9.3.a, ``Steam Generator Tube 
Surveillance Program, Inspection Frequencies.'' Specifically, the 
proposed changes revise the Farley Nuclear Plant, Unit 1 TS to allow a 
40-month inspection interval after its first (post-replacement) 
inservice inspection, rather than after two consecutive inspections 
resulting in C-1 classification.
    Date of issuance: September 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 157.
    Facility Operating License No. NPF-2: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53991). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 20, 2002.

[[Page 63705]]

    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: September 20, 2001, as 
supplemented by letters dated January 24, April 25, July 3, and July 
16, 2002.
    Brief description of amendments: The amendments revise the 
Technical Specifications to support extension of certain surveillance 
requirements from ``92 days'' to ``92 days on an alternate test 
basis.''
    Date of issuance: September 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 234 and 176.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59514). The supplements dated January 24, April 25, July 3, and July 
16, 2002, provided clarifying information that did not change the scope 
of the September 20, 2001, application nor the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 26, 2002.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: August 2, 2001, as supplemented by 
letters dated March 6, April 2, and June 25, 2002. The supplemental 
information provided clarification that did not change the scope or the 
initial no significant hazards consideration determination.
    Brief description of amendments: The amendments revise the 
Technical Specification permitting a one time extension of Title 10 of 
the Code of Federal Regulations, Part 50, Appendix J, Option B, 
Performance-Based Leakage-Test Requirements.
    Date of issuance: September 17, 2002.
    Effective date: September 17, 2002.
    Amendment Nos.: Unit 1-143; Unit 2-131.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50959). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 4, 2002 (TSC 00-04).
    Description of amendment request: The amendments relocated certain 
Technical Specification (TS) surveillance requirements to the Sequoyah 
Technical Requirements Manual.
    Date of issuance: September 5, 2002.
    Effective date: September 5, 2002.
    Amendment Nos: 277 and 268.
    Facility Operating License No. DPR-79: Amendments revise the TSs.
    Date of initial notice in Federal Register:April 16, 2002 (67 FR 
18648). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 21, 2001, as 
supplemented by letters dated June 11, July 19, August 9 and 30, and 
September 5 and 12, 2002 (TS 00-06).
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to allow irradiation of up to 2256 
tritium-producing burnable absorber rods.
    Date of issuance: September 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
prior to irradition of TPBARs.
    Amendment Nos.: Unit 1-278, Unit 2-269.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: December 17, 2001 (66 
FR 65000). The supplemental letters provided clarifying information 
that did not expand the application beyond the scope of the initial 
notice and did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in an Environmental Assessment dated September 23, 2002 (67 FR 59581) 
and in a Safety Evaluation dated September 30, 2002.
    No significant hazards consideration comments received: Comments 
were received in response to the staff's proposed no significant 
hazards consideration determination that was published in the December 
17, 2001, Federal Register, from Dr. Kenneth D. Bergeron and The Blue 
Ridge Environmental Defense League (BREDL). BREDL's comments 
incorporated Mr. Bergeron's comments by reference. These comments were 
addressed by the staff in a letter from Dr. Brian Sheron to Mr. 
Bergeron dated September 6, 2002, with a copy to BREDL (Accession No. 
ML022410310).

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 12, 2001, as 
supplemented September 17, 2002 (TS 01-04).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah Nuclear Plant, Units 1 and 2 Technical 
Specification (TS) 3/4 6.5.1 and associated Bases to reflect an 
increase in the ice condenser basket weight from 1071 pounds to 1145 
pounds and the total ice condenser ice weight from 2,082,024 pounds to 
2,225,880 pounds. This change is being made in response to a reanalysis 
by Westinghouse Electric Company that identified a modeling input error 
used in the original analysis.
    Date of issuance: September 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 279 & 270.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
934). The September 17, 2002, supplement contained clarifying 
information only, and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 30, 2002.
    No significant hazards consideration comments received: No.

[[Page 63706]]

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: August 20, 2001, as supplemented 
by letters of October 29, November 14, November 21, December 7, 
December 19, 2001, and January 14, February 19, February 21, May 21, 
May 23, and July 30, 2002.
    Brief description of amendment: The amendment allows Watts Bar 
Nuclear Plant, Unit 1, to irradiate up to 2304 tritium-producing 
burnable absorber rods in the reactor core each fuel cycle.
    Date of issuance: September 23, 2002.
    Effective date: As of the date of issuance and shall be implemented 
prior to starting up from the outage where TVA inserts tritium-
producing burnable absorber rods in the core.
    Amendment No.: 40.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 2001 (66 
FR 65005). The supplemental letters provided clarifying information 
that was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in an Environmental Assessment dated August 20, 2002 (ADAMS Accession 
No. ML022320905) and in a Safety Evaluation dated September 23, 2002.
    No significant hazards consideration comments received: Comments 
were received in response to the staff's proposed no significant 
hazards consideration determination (66 FR 65005) from Dr. Kenneth D. 
Bergeron and The Blue Ridge Environmental Defense League (BREDL). 
BREDL's comments incorporated Dr. Bergeron's comments by reference. 
These comments were addressed by the staff in a letter from Dr. Brian 
Sheron to Dr. Bergeron dated September 6, 2002, with a copy to BREDL 
(Accession No. ML022410310). The staff made a final determination that 
the amendment involves no significant hazards consideration, which is 
contained in the Safety Evaluation dated September 23, 2002.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: October 23, 2001, as supplemented by 
letters dated July 23, August 29, and September 6, 2002.
    Brief description of amendments: The amendments revise TS 5.5.9, 
``Steam Generator Tube Inspection Report,'' to permit installation of 
leak-tight sleeves in the Comanche Peak Steam Electric Station, Unit 1, 
steam generators as an alternative to plugging defective steam 
generator tubes.
    Date of issuance: September 25, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 101.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66473). The July 23, August 29, and September 6, 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 amendments is contained 
in a Safety Evaluation dated September 25, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 17, 2002.
    Brief description of amendment: The amendment revises Limiting 
Conditions for Operation (LCOs), Required Actions for LCOs, 
Surveillance Requirements, and Tables specifying requirements on 
instrumentation in the following Technical Specifications: (1) TS 
3.3.6, ``Containment Purge Isolation Instrumentation''; (2) TS 3.3.7, 
``Control Room Emergency Ventilation System (CREVS) Instrumentation''; 
(3) TS 3.3.8, ``Emergency Exhaust System (EES) Actuation 
Instrumentation''; and (4) TS 3.9.4, ``Containment Penetrations.'' The 
revisions allow the equipment hatch and the emergency air lock to be 
open in refueling outages during core alterations and/or movement of 
irradiated fuel within containment.
    Date of issuance: September 9, 2002.
    Effective date: September 9, 2002, and shall be implemented, 
including the incorporation of the changes to the Bases of the 
Technical Specifications and to the Final Safety Analysis Report for 
Callaway, as described in the licensee's letter of June 17, 2002, prior 
to entry into Mode 6 during Refueling Outage 12 that is scheduled for 
the Fall of 2002.
    Amendment No.: 152.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 23, 2002 (67 FR 
48222). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 9, 2002.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: February 26, 2002, as 
supplemented by letter dated July 15, 2002.
    Brief description of amendment: These amendments revise the 
surveillance frequency of the quench and recirculation spray system 
nozzles, from a time period of every 10 years to whenever maintenance 
is conducted that could contribute to nozzle blockage.
    Date of issuance: October 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 233 and 215.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21296). The supplemental letter dated July 15, 2002, provided 
clarifying information that did not change the scope of the February 
26, 2002, application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 4th 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. 02-25990 Filed 10-11-02; 8:45 am]
BILLING CODE 7590-01-P