[Federal Register Volume 67, Number 180 (Tuesday, September 17, 2002)]
[Notices]
[Pages 58635-58653]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-23358]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Application 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, August 23, 2002, through September 5, 2002. 
The last biweekly notice was published on September 3, 2002 (67 FR 
56317).

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

[[Page 58636]]

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 October 17, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714,\1\ which is 
available at the Commission's PDR, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
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    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest .
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
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    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.

[[Page 58637]]

    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].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: July 31, 2002.
    Description of amendment request: The proposed amendment would add 
a Surveillance Requirement (SR) to Technical Specification 3.2.2, 
``Minimum Critical Power Ratio (MCPR),'' that requires determination of 
the MCPR limits following completion of control rod scram time testing. 
The proposed SR would provide for the required evaluation necessary to 
apply faster scram times to provide for improved MCPR 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds a new surveillance requirement (SR) to 
the Minimum Critical Power Ratio (MCPR) Technical Specification (TS) 
which requires determination of the MCPR operating limit following 
the completion of scram time testing of the control rods. Use of the 
scram speed in determining the MCPR operating limit (i.e., Option B) 
is an alternative to the current method for determining the 
operating limit (i.e., Option A). The probability of an accident 
previously evaluated is unrelated to the MCPR operating limit that 
is provided to ensure no fuel damage results during anticipated 
operational occurrences. This is an operational limit to ensure 
conditions following an assumed accident do not result in fuel 
failure and therefore do not contribute to the occurrence of an 
accident. No active or passive failure mechanisms that could lead to 
an accident are affected by this proposed change.
    The consequences of a previously evaluated accident are not 
significantly increased. The proposed change ensures that the 
appropriate operating limit is in place. By implementing the correct 
operating limit the safety limit will continue to be ensured. 
Ensuring the safety limit is not exceeded will result in prevention 
of fuel failure. Therefore, since there is no increase in the 
potential for fuel failure there is no increase in the consequences 
of any accidents previously evaluated.
    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 or a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The addition of a new SR to the MCPR TS does not involve the use 
or installation of new equipment. Installed equipment is not 
operated in a new or different manner. No new or different system 
interactions are created, and no new processes are introduced. No 
new failures have been created by the addition of the proposed SR 
and the use of the alternate method for determining the MCPR 
operating limit.
    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.
    Use of Option B for determining the MCPR operating limit will 
result in a reduced operating limit in comparison to the use of 
Option A. However, a reduction in the operating limit margin does 
not result in a reduction in the safety margin. The MCPR safety 
limit remains the same regardless of the method used for determining 
the operating limit. All analyzed transient results remain well 
within the design values for structure, systems, and components.
    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: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of amendment request: August 6, 2002.
    Description of amendment request: The proposed amendment would 
allow the installation of up to four lead fuel assemblies (LFAs) 
manufactured by Framatome ANP, Inc. (FRA-ANP) into the Unit 2 Cycles 15 
and 16 cores. Currently, Technical Specification 4.2.1, Fuel 
Assemblies, only allows fuel that is clad with either zircaloy or 
ZIRLO. The FRA-ANP LFA utilizes M5\TM\ alloy for the fuel cladding.
    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, 10 CFR 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 Framatome ANP, Inc. (FRA-ANP) 
fuel assemblies into Calvert Cliffs Unit 2 that have fuel rods clad 
in an alloy that does 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 material, 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. The addition 
of what will be an approved temporary exemption for Unit 2 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 zircaloy 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. The similarity of ZIRLO 
to zircaloy was previously approved by the

[[Page 58638]]

Nuclear Regulatory Commission. Therefore, the addition of the 
advanced cladding M5\TM\ 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 zircaloy 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 cores. Therefore, the 
addition of the advanced cladding M5\TM\ 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 FRA-ANP LFAs. The assemblies use the advanced cladding 
material M5\TM\ that is 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.
    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.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 1, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.7.1.1, ``Plant Systems: Turbine Cycle 
Safety Valves,'' to reflect results of a reanalysis of 
overpressurization events to reinstate the capability to operate, at 
corresponding reduced power levels, with up to four main steam line 
code safety valves in each main steam line inoperable.
    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 change will revise Specification 3.7.1.1 in 
accordance with revised overpressurization analyses to reinstate the 
capability to operate at corresponding reduced power levels with up 
to four main steam line code safety valves (MSSVs) in each main 
steam line inoperable. The MSSVs ensure the American Society of 
Mechanical Engineers (ASME) Code, Section III requirements are 
maintained to limit secondary system pressure to within 110 percent 
of design pressure during the most severe anticipated system 
operational transient. Operation with less than the full number of 
MSSVs is permitted as long as thermal power is restricted (and the 
Power Level-High trip setpoint is reset within the specified 
timeframe). These actions place restrictions on the allowable 
thermal power so that the energy transfer to the most limiting steam 
generator (SG) is not greater than the available relief capacity for 
that generator.
    These changes are consistent with the Unit No. 2 Final Safety 
Analysis Report (FSAR) design description and analysis assumptions 
where the MSSVs provide the required overpressure protection. The 
proposed change provides assurance that the secondary side pressure 
remains within the bounds of the safety analyses; therefore, the 
proposed change will 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 ensures that adequate secondary side 
overpressure protection is available and properly maintained. This 
change limits plant power level based on the number of operable 
MSSVs. The actions require a reduction in power when the number of 
MSSVs is less than the full complement for each SG and also required 
a reduction in the Power Level-High trip setpoint.
    The proposed change does not involve a physical alteration of 
the plant or change the plant configuration (no new or different 
type of equipment will be installed). The proposed change only 
reinstates a previously authorized mode of operation based upon 
revised analyses. It does not require any new or unusual operator 
actions. The change does not alter the way any structure, system, or 
component functions and does not alter the manner in which the plant 
is operated. The change does not introduce any new failure modes. 
Therefore, the proposed change will 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 MSSVs ensure the ASME Code, Section III requirements are 
maintained to limit the secondary system pressure to within 110 
percent of the design pressure. This ensures that the overpressure 
protection system can cope with all operational and transient 
events. Plant operation with a reduced number of MSSVs is subject to 
the same considerations as the condition when all MSSVs are 
operable, i.e., a transient overpressure event must not exceed the 
acceptance criteria specified in the Unit No. 2 FSAR. Restricting 
the thermal power provides this assurance. Reducing the Power Level-
High trip setpoint (within the specified timeframe provides 
additional assurance).
    These actions place restrictions on the allowable thermal power 
so that the energy transfer to the most limiting SG is not greater 
than the available relief capacity for that generator, consequently 
these actions ensure the margin of safety is maintained consistent 
with the analysis bases.
    The proposed change does not impact any acceptance criteria for 
the design basis accidents described in the FSAR and does not impact 
the consequences of accidents previously evaluated. The proposed 
change provides assurance that the secondary side pressure remains 
within the bounds of the safety analyses; therefore, the proposed 
change will not involve a significant increase in the probability or 
consequences of an accident previously evaluated. Therefore, the 
proposed change will not result in a reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jacob I. Zimmerman, Acting.

[[Page 58639]]

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 7, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 6.9.1.8, ``Core Operating Limits 
Report,'' to update the list of documents that describe the analytical 
methods used to determine the 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to delete the document contained in section 
6.9.1.8b.4 is required since it has been superceded by the most 
recent methodology as described in the document contained in section 
6.9.1.8b.15 (renumbered 6.9.1.8b.14). Adding the new document 
associated with the new section 6.9.1.8b.15 to the list of 
references is required for completeness. This change has no impact 
on plant equipment operation. Since the changes only affect 
description of the safety analysis methodology and do not revise any 
setpoints assumed in the accident analyses, they cannot affect the 
likelihood or consequences of accidents. Therefore, this change 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 will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. These changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will 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 have no impact on plant equipment 
operation. The proposed changes do not revise any setpoints assumed 
in the analyses and do not affect the acceptance criteria for the 
Steam Line Break accident. Therefore, the proposed changes will not 
result in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jacob I. Zimmerman, Acting.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 12, 2002.
    Description of amendment request: The proposed amendment would 
change the surveillance requirements for the emergency diesel 
generators (EDGs) in Technical Specification (TS) 3/4.8.1.1, 
``Electrical Power Systems--A.C. Sources--Operating'' and TS 3/4.8.1.2, 
``Electrical Power Systems--Shutdown.'' In addition, TS Section 6.0, 
``Administrative Controls,'' would be revised to add a new TS to define 
the program requirements for testing the EDG fuel oil. The TS index and 
the TS Bases would also be revised to address 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 Technical Specification changes associated with 
revising the surveillance requirements for the Millstone Unit No. 2 
emergency diesel generators and adding a new specification to define 
the program requirements for testing of the emergency diesel 
generator fuel oil will not cause an accident to occur and will not 
result in any change in the operation of the associated accident 
mitigation equipment. The ability of the equipment associated with 
the proposed changes to mitigate the design basis accidents will not 
be affected. The proposed Technical Specification surveillance 
requirements are sufficient to ensure the required accident 
mitigation equipment will be available and function properly for 
design basis accident mitigation. In addition, the design basis 
accidents will remain the same postulated events described in the 
Millstone Unit No. 2 Final Safety Analysis Report, and the 
consequences of those events will not be affected. Therefore, the 
proposed changes will not increase the probability or consequences 
of an accident previously evaluated.
    The additional proposed changes to the Technical Specifications 
(e.g., renumbering a requirement, modifying an index page, 
relocating a footnote requirement, relocating requirements to 
surveillance notes, relocating part of a surveillance requirement to 
be a separate surveillance requirement, clarifying the EDGs loads 
required to be energized for at least 5 minutes, clarifying the EDGs 
loads that should remain energized by offsite power) will not result 
in any technical changes to the current requirements. Therefore, 
these additional 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 to the Technical Specifications do not 
impact any system or component that could cause an accident. The 
proposed changes will not alter the plant configuration (no new or 
different type of equipment will be installed) or require any 
unusual operator actions. The proposed changes will not alter the 
way any structure, system, or component functions, and will not 
alter the manner in which the plant is operated. There will be no 
adverse effect on plant operation or accident mitigation equipment. 
The response of the plant and the operators following an accident 
will not be different. In addition, the proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will 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 Technical Specification changes associated with 
revising the surveillance requirements for the Millstone Unit No. 2 
emergency diesel generators and adding a new specification to define 
the program requirements for testing of the emergency diesel 
generator fuel oil will not cause an accident to occur and will not 
result in any change in the operation of the associated accident 
mitigation equipment. The equipment associatedwith the proposed 
Technical Specification changes will continue to be able to mitigate 
the design basis accidents as assumed in the safety analysis. The 
proposed surveillance requirements are adequate to ensure proper 
operation of the affected accident mitigation equipment. In 
addition, the proposed changes will not affect equipment design or 
operation, and there are no changes being made to the Technical 
Specification required safety limits or safety system settings. The 
proposed Technical Specification changes will provide adequate 
control measures to ensure the accident mitigation functions are 
maintained. Therefore, the proposed changes will not result in a 
reduction in a margin of safety.
    The additional proposed changes to the Technical Specifications 
(e.g., renumbering a requirement, modifying an index page, 
relocating a footnote requirement, relocating requirements to 
surveillance notes, relocating part of a surveillance requirement to 
be a separate surveillance requirement, clarifying the EDGs loads 
required to be energized for at least 5 minutes, clarifying the EDGs 
loads that should remain energized by offsite power) will not result 
in any technical changes to the current requirements. Therefore, 
these additional changes will not result in a reduction in a margin 
of safety.


[[Page 58640]]


    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.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 14, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications related to reactivity control 
systems, power distribution limits, and special test exceptions. The 
purpose of the proposed changes are to remove ambiguity and improve 
usability of the current Technical Specifications.
    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 Technical Specification changes associated with the 
deletion of special test exceptions in Specifications 3/4.10.3, 3/
4.10.4 and 3/4.10.5, changes to reflect the current Millstone Unit 
No. 2 design (i.e. full length CEAs [control element assemblies]), 
changes that limit the Mode applicability requirement for Shutdown 
Margin requirements (Specifications 3/4.1.1.1 and 3/4.1.1.2), and 
changes to action requirements and surveillance requirements will 
not cause an accident to occur and will not result in any change in 
operation of the mitigation equipment. The proposed changes in 
Specification 3/4.1.3.1 have no effect on the operability and 
alignment ofCEAs. The proposed allowed outage times and shutdown 
times are reasonable and consistent with the industry guidelines to 
ensure the accident mitigation equipment will be restored in a 
timely manner. In addition the design basis accident will remain the 
same postulated events described in the Millstone Unit No. 2 Final 
Safety Analysis Report. Since the initial conditions and assumptions 
included in the safety analyses are unchanged, the consequences of 
the postulated events remain unchanged. Therefore the proposed 
changes will not increase the probability or consequences of an 
accident previously evaluated.
    The additional proposed changes to the Technical Specifications 
(e.g.[,] combining requirements, re-ordering requirements, 
relocating information to the Bases, modifying index pages, deletion 
or addition of footnotes) will not result in any technical changes 
to the current requirements. Therefore, these additional changes 
will not increase the probability or consequences of an accident 
previously evaluate[d].
    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. Since the requirements remain the 
same, the proposed changes do not alter the way any system, 
structure, 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 will 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 assumptions. The 
proposed changes do not decrease the scope of equipment currently 
required to operate 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: Victor Nerses, Acting.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 14, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) related to Containment 
Systems. Specifically, the proposed changes would: (1) Add a new 
requirement for a Containment Tendon Surveillance Program to TS Section 
6.0, ``Administrative Controls;'' (2) delete TS 3/4.6.1.6, 
``Containment Structural Integrity;'' (3) revise TS 3/4.6.1.1, 
``Containment Integrity,'' to add a new surveillance requirement that 
would require that containment structural integrity be verified in 
accordance with the Containment Tendon Surveillance Program; (4) revise 
TS 3/4.6.3.1, ``Containment Isolation Valves'' to add a new action 
statement that would increase the allowed outage time (AOT) from 4 
hours to 72 hours for Containment Isolation Valves (CIVs) in closed 
systems; (5) make other changes to the TSs for Containment Integrity 
and CIVs to provide clarity to the TSs; and (6) make other 
administrative type changes. In addition, the TS Bases would be revised 
to address 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 Technical Specification changes associated with 
both containment integrity and CIVs that will remove ambiguity, 
improve usability, and increase AOT for CIVs in closed systems, will 
not cause an accident to occur. Operablity requirements for 
containment integrity and CIVs will remain the same. The ability of 
the equipment associated with the proposed changes to mitigate the 
design basis accidents will not be affected. The proposed Technical 
Specification requirements are sufficient to ensure the required 
accident mitigation equipment will be available and function 
properly for design basis accident mitigation. The proposed allowed 
outage time is reasonable and consistent with standard industry 
guidelines to ensure the accident mitigation equipment will be 
restored in a timely manner. In addition, the design basis accidents 
will remain the same postulated events described in the Millstone 
Unit No. 2 Final Safety Analysis Report, and the consequences of 
those events will not be affected. Therefore, the proposed changes 
will not increase the probability or consequences of an accident 
previously evaluated.
    The additional proposed changes to the Technical Specifications 
(e.g., changes to index, renumbering a requirement) will not result 
in any technical changes to the current requirements. Therefore, 
these additional 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.

[[Page 58641]]

    The proposed changes to the Technical Specifications do not 
impact any system or component that could cause an accident. The 
proposed changes will not alter the plant configuration (no new or 
different type of equipment will be installed) or require any 
unusual operator actions. The proposed changes will not alter the 
way any structure, system, or component functions, and will not 
alter the manner in which the plant is operated. The response of the 
plant and the operators following an accident will not be different. 
In addition, the proposed changes do not introduce any new failure 
modes. Therefore, the proposed changes will 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 Technical Specification changes associated with 
both containment integrity and CIVs that will remove ambiguity, 
improve usability, and increase AOT for CIVs in closed systems, will 
not cause an accident to occur. Operablity requirements for 
containment integrity and CIVs will remain the same. Although, 
Containment Structural Integrity and Containment Integrity 
Specifications are combined, operability of the containment 
structure will continue to be maintained as part of a surveillance 
program. The equipment associated with the proposed Technical 
Specification changes will continue to be able to mitigate the 
design basis accidents as assumed in the safety analysis. The 
proposed allowed outage time is reasonable and consistent with 
standard industry guidelines to ensure the accident mitigation 
equipment will be restored in a timely manner. In addition, the 
proposed changes will not affect equipment design or operation, and 
there are no changes being made to the Technical Specification 
required safety limits or safety system settings. The proposed 
Technical Specification changes will provide adequate control 
measures to ensure the accident mitigation functions are maintained. 
Therefore, the proposed changes will not result in a reduction in a 
margin of safety.
    The additional proposed changes to the Technical Specifications 
(e.g., changes to index, renumbering a requirement) will not result 
in any technical changes to the current requirements. Therefore, 
these additional changes will not result in a reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: August 22, 2002.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.8.1 to allow a one-time extension 
of the completion times for each Keowee Hydro Unit (KHU). This would 
accommodate a complete inspection and overhaul of each KHU that is 
expected to take more time than the current TS 3.8.1 completion time 
would allow.
    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[.]
    No. The change involves an extension of the Completion Times for 
TS 3.8.1 Required Action C.2.2.5 and Required Action H.2. During the 
time that one KHU is inoperable for 72 hours or both KHUs 
are inoperable, a LCT [Lee combustion turbine] will be energizing 
both standby buses, two available offsite power sources will be 
maintained available, and maintenance on electrical distribution 
systems will not be performed unless necessary. Extending the 
Completion Times will decrease the likelihood of an unplanned forced 
shutdown of all three Oconee Units and the potential safety 
consequences and operational risks associated with that action. 
Avoiding this risk offsets the risks associated with having a design 
basis event during the additional completion time for having one or 
both KHUs inoperable.
    Extending the Completion Time does not involve: (1) A physical 
alteration to the Oconee Units; (2) the installation of new or 
different equipment; (3) operating any installed equipment in a new 
or different manner; or (4) a change to any set points for 
parameters which initiative protective or mitigation action.
    There is no adverse impact on containment integrity, 
radiological release pathways, fuel design, filtration systems, main 
steam relief valve set points, or radwaste systems. No new 
radiological release pathways are created.
    The consequences of an event occurring during the extended 
Completion Time are the same as those that would occur during the 
existing Completion Time. A risk assessment shows that the 
additional time coupled with compensatory measures results in an 
acceptable level of risk.
    Therefore, the probability or consequences of an accident 
previously evaluated is not significantly increased.
    2. Create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated[.]
    No. This change involves an extension of the Completion Times 
for TS 3.8.1 Required Actions C.2.2.5 and H.2 associated with 
restoring compliance with TS LCO 3.8.1.C. During the time period 
that both KHUs are inoperable, the safety function for the emergency 
power source will be fulfilled by the LCTs. Compensatory measures 
previously specified will be in place.
    Extending the Completion Times does not involve a physical 
effect on the unit, nor is there any increased risk of a unit trip 
or reactivity excursion. No new failure modes or credible accident 
scenarios are postulated from this activity.
    Therefore, the possibility of a new or different kind of 
accident from any kind of accident previously evaluated is not 
created.
    3. Involve a significant reduction in a margin of safety.
    No. This change involves an extension of the Completion Times 
for TS 3.8.1 Required Actions C.2.2.5 and H.2 associated with 
restoring compliance with TS LCO 3.8.1.C. During the time period 
that both KHUs are inoperable, the safety function for the emergency 
power source will be fulfilled by the LCTs. Compensatory measures 
previously specified will be in place to minimize electrical power 
system vulnerabilities.
    Extending the Completion Time does not involve: (1) A physical 
alteration of the Oconee Units; (2) the installation of new or 
different equipment; (3) operating any installed equipment in a new 
or different manner; (4) a change to any set points for parameters 
which initiate protective or mitigation action; or (5) any impact on 
the fission product barriers or safety limits.
    Therefore, this request 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

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: July 10, 2002.
    Description of amendment request: The proposed change would 
increase the control rod scram time testing interval from 120 days to 
200 days of full 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

[[Page 58642]]

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 will not adversely impact plant operation. 
There will be no change in the method of performing the tests. The 
extended test frequency will provide some positive safety benefits 
by reducing the complexity of half of the control rod sequence 
exchange maneuvers, reducing the likelihood of a reactivity or fuel 
related event.
    The actual rod insertion times and control rod reliability are 
not impacted by this proposed change; only the probability of 
detecting slow rods is impacted. The potential consequence of the 
proposed change is that one or more slow rods that would have been 
detected under the current 120-day frequency, may not be detected 
due to a reduced number of tests under the 200-day frequency.
    Historical data shows that the River Bend Station control rod 
insertion frequency is highly reliable and rod insertion tests meet 
the scram time limits 99.949% of the time. Statistical analysis also 
demonstrates that the extended frequency would have little impact on 
the ability to detect slow rods in the sampling tests.
    There is no safety consequence resulting from ``slow'' rods so 
long as the plant does not exceed the Technical Specification 3.1.4 
Limiting Condition for Operation requirement of no more than 10 slow 
rods in the entire core or no two OPERABLE ``slow'' rods occupying 
adjacent positions. It is highly unlikely that a combination of 
missed detections and known ``slow'' rods would lead to the 
requirement to take action in accordance with Technical 
Specification 3.1.4. as discussed in the supporting analysis. 
Therefore, it is highly unlikely that the reduction in test 
frequency would have any impact on plant operation or safety.
    The plant safety analysis assumes that all 10 slow rods take 7 
seconds to reach notch position 13 which is very conservative based 
on actual rod performance. Control rod data shows that rods that 
have failed the time requirements are usually only a fraction of a 
second slower. The low probability of MODE 1 operation with excess 
slow rods combined with the historically low incidence of failure, 
leads to the conclusion that the probability or consequences of 
accidents previously evaluated are not significantly increased.
    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.
    The proposed change will make no change to plant configuration 
or test procedures. The proposed change does not impact the 
operation of the plant except to reduce the number of required tests 
and slightly increase the probability of failing to detect a slow 
control rod. Operating with possibly one or two undetected slow rods 
does not create the possibility of an accident, since sudden control 
rod insertion by scram is an accident mitigation action.
    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 River Bend Station accident analyses assume a certain 
negative reactivity time function associated with scrams. So long as 
the Limiting Condition for Operation of Technical Specification 
3.1.4 is met, that is, there are no more than 10 slow control rods 
in the entire core or two operable ``slow'' rods occupying adjacent 
locations, all accident analysis assumptions are met and there is no 
reduction in any margin of safety. The proposed change does not 
impact the Technical Specification Limiting Condition for Operation 
or any other allowable operating condition. The potential for an 
increase in the probability of being outside acceptable operating 
conditions due to this proposed change is insignificant. 
Calculations have demonstrated that the likelihood of detecting four 
slow rods with proposed testing frequency over a fuel cycle is lower 
than that with the current testing frequency by a negligible amount. 
The difference is even smaller for detecting a greater number of 
slow rods over a cycle. Therefore, since there is no impact on 
allowable operating parameters and the likelihood of detecting 
significant numbers of slow rods is only negligibly affected, there 
is no significant reduction in a margin of safety.
    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket No. 50-10, Dresden Nuclear Power 
Station (DNPS), Unit 1, Grundy County, Illinois

    Date of amendment request: August 1, 2002.
    Description of amendment request: The proposed changes revise the 
Operating License to update references to plant documents and delete 
Technical Specification limiting conditions for required equipment and 
surveillance requirements that no longer apply or are being relocated 
to the DNPS Technical Requirements Manual. In addition, the proposed 
changes delete or revise administrative control and staffing 
requirements that either no longer apply or have changed due to the 
Unit 1 Fuel Storage Pool no longer containing irradiated fuel 
assemblies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The accidents previously evaluated in the Defueled Safety 
Analysis Report (DSAR) affecting nuclear safety only involve the 
storage and handling of irradiated fuel. In each analyzed accident, 
irradiated fuel is assumed to be stored in the Dresden Nuclear Power 
Station (DNPS), Unit 1 Fuel Storage Pool. Since irradiated fuel has 
been permanently removed from the Unit 1 Fuel Storage Pool, the 
previously analyzed accidents are no longer credible, and therefore 
can not possibly occur. The proposed Technical Specifications (TS) 
changes delete requirements involving storage and handling of 
irradiated fuel, sealed source contamination, liquid radwaste 
storage radioactivity, written procedures, the Process Control 
Program and the unit staff, and reassign plant management 
responsibilities. The proposed Amended Facility Operating License 
(OL) changes are administrative in nature in that they only correct 
references to superseded plant documents. Based on the above, the 
proposed OL and TS changes 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.
    The proposed changes delete requirements involving storage and 
handling of irradiated fuel, sealed source contamination, liquid 
radwaste storage radioactivity, written procedures, the unit staff 
and the Process Control Program; and reassign plant management 
responsibilities. Deletion of requirements involving storage and 
handling of irradiated fuel is consistent with the current plant 
configuration with irradiated fuel permanently removed from the Unit 
1 Fuel Storage Pool and stored in either the ISFSI [independent 
spent fuel storage installation] or the Unit 3 Spent Fuel Pool. 
Irradiated fuel in the ISFSI is controlled in accordance with 10 CFR 
[part] 72, ``Licensing Requirements for the Independent Storage of 
Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-
Related Greater Than Class C Waste.'' Irradiated fuel in the Unit 3 
Spent Fuel Pool is controlled by the DNPS Units 2 and 3 TS in 
accordance with 10 CFR [part] 50, ``Domestic Licensing of Production 
and Utilization Facilities.'' Since accident

[[Page 58643]]

analysis for Unit 1 irradiated fuel is now controlled by either 10 
CFR [part 72 or the DNPS Units 2 and 3 TS, the deletion of DNPS Unit 
1 TS requirements involving storage and handling of irradiated fuel 
will not create new or different kinds of accidents. Relocation of 
requirements for liquid radwaste storage radioactivity and sealed 
source contamination will not create new or different kinds of 
accident[s] since the requirements will still be applicable, but 
specified in the DNPS Technical Requirements Manual (TRM) not the 
DNPS Unit 1 TS. Similarly, Process Control Program requirements are 
redundantly contained in the DNPS Units 2 and 3 Updated Final Safety 
Analysis Report (UFSAR). Therefore, deletion of requirements for the 
Process Control Program will not contribute to the creation of a new 
or different kind of accident from any accident previously 
evaluated. Deletion of requirements for written procedures and the 
unit staff, and reassignment of plant management responsibilities 
are administrative changes only and will not contribute to the 
creation of a new or different kind of accident from any accident 
previously evaluated. In addition, the proposed OL changes are also 
administrative in nature and will not contribute to the creation of 
a new or different kind of accident from any accident previously 
evaluated. The proposed changes do not physically alter the plant 
and will not alter the operation of the structures, systems, and 
components as described in the DSAR. Therefore, a new or different 
kind of accident from any accident previously evaluated will not be 
created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The removal of TS requirements involving storage and handling of 
irradiated fuel only corrects the TS to conform to the current plant 
conditions (i.e., irradiated fuel permanently removed from the Fuel 
Storage Pool). Unit 1 irradiated fuel storage and handling is now 
controlled in accordance with either 10 CFR [part] 72 or the DNPS 
Units 2 and 3 TS (required by 10 CFR [part] 50), not the current 
DNPS Unit 1 TS. Thus, any changes to the DNPS Unit 1 TS involving 
storage and handling of irradiated fuel do not reduce any margin of 
safety. The relocation of the sealed source contamination and liquid 
storage radioactivity requirements from the DNPS Unit 1 TS to the 
DNPS TRM does not reduce any safety margin since the requirements 
still pertain. Process Control Program requirements are redundantly 
contained in the DNPS Units 2 and 3 UFSAR. Therefore, deletion of 
Process Control Program requirements from the DNPS Unit 1 TS does 
not reduce any safety margin since UFSAR changes are controlled 
under the provisions of 10 CFR 50.59, ``Changes, tests, and 
experiments.'' The deletion of written procedure requirements and 
unit staff requirements, and reassignment of plant management 
responsibilities are administrative changes only. In addition, the 
proposed OL changes are also administrative in nature in that they 
only correct references to obsolete plant documents. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

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

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

    Date of amendment request: May 31, 2002.
    Description of amendment request: The proposed amendment would 
provide specific actions and increase restoration time for an 
inoperable battery charger; relocate preventative maintenance 
surveillance requirements for the battery charger from the Technical 
Specifications (TSs) to the Technical Requirements Manual (TRM); 
replace battery specific gravity monitoring with battery float 
monitoring; relocate battery float voltage and battery cell voltage, 
level, and temperature from the TSs to the TRM, and revise the 
associated surveillance requirements; create a new battery monitoring 
and maintenance program; provide specific actions with increased 
restoration time for certain battery and battery cell parameter out-of-
limits conditions; eliminate the once per 60-month restriction on 
crediting performance discharge test for service test and restrict its 
use to the modified performance discharge test; revise the duration of 
the battery charger service test from 8 hours to 4 hours; revise the 
frequency of the battery performance discharge test; and delete 
surveillance requirements that provide excessive detail.
    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 class 1E direct current (DC) electrical power system 
including associated battery chargers are not initiators to any 
accident sequence analyzed in the Updated Final Safety Analysis 
Report (UFSAR). Operation in accordance with the proposed Technical 
Specification (TS) ensures that the DC system is capable of 
performing its function as described in the UFSAR, therefore the 
mitigative functions supported by the DC system will continue to 
provide the protection assumed by the analysis. The relocation of 
preventive maintenance surveillances, certain operating limits and 
actions to either the Technical Requirements Manual (TRM), TS Bases, 
or newly-created TS 6.8.4.h, ``Battery Monitoring and Maintenance 
Program,'' will not challenge the ability of the DC system to 
perform its design function. Appropriate monitoring and maintenance, 
consistent with industry standards, will continue to be performed. 
In addition, the DC system is within the scope of 10 CFR 50.65, 
``Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants,'' which will ensure the control of maintenance 
activities associated with the DC system.
    These changes do not involve any physical change to structures, 
systems, or components (SSCs) and do not alter the method of 
operation or control of SSCs. The current assumptions in the safety 
analysis regarding accident initiators and mitigation of accidents 
are unaffected by these changes. No additional failure modes or 
mechanisms are being introduced and the likelihood of previously 
analyzed failures remains unchanged.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the UFSAR will not be 
affected by these changes. Therefore, the consequences of previously 
analyzed accidents will not increase because of these changes.
    Based on the above discussion, the proposed TS changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
plant. No new equipment is being introduced, and installed equipment 
is not being operated in a new or different manner. There are no 
setpoints, at which protective or mitigative actions are initiated, 
affected by this change. These changes will not alter the manner in 
which equipment operation is initiated, nor will the function 
demands on credited equipment be changed. Any alteration in 
procedures will continue to ensure that the plant remains within 
analyzed limits, and no change is being made to the procedures 
relied upon to respond to an off-normal event as described in the 
UFSAR. As such, no new failure modes are being introduced. The 
changes do not alter assumptions made in the safety analysis and 
licensing basis.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The margin of safety is established through equipment design, 
operating parameters, and

[[Page 58644]]

the setpoints at which automatic actions are initiated. The proposed 
changes are acceptable because the operability of the DC system is 
unaffected, there is no detrimental impact on any equipment design 
parameter, and the plant will still be required to operate within 
assumed conditions. Operation in accordance with the proposed TS 
ensures that the DC system is capable of performing its function as 
described in the UFSAR; therefore, the support of the DC system to 
the plant response to analyzed events will continue to provide the 
margins of safety assumed by the analysis. The relocation of 
preventive maintenance surveillances, certain operating limits and 
actions to either the TRM, TS Bases, or newly-created TS 6.8.4.h, 
``Battery Monitoring and Maintenance Program,'' will not challenge 
the ability of the DC system to perform its design function. 
Appropriate monitoring and maintenance, consistent with industry 
standards, will continue to be performed. In addition, the DC system 
is within the scope of 10 CFR 50.65, ``Requirements for monitoring 
the effectiveness of maintenance at nuclear power plants,'' which 
will ensure the control of maintenance activities associated with 
the DC system. This provides sufficient management control of the 
requirements that assure the batteries are maintained in a highly 
reliable condition.
    The increased restoration times and revised criteria for 
monitoring the capacity of the battery and battery chargers to 
perform their intended function, are reasonable and generally 
consistent with approved standards, guidance and regulations.
    Based on the above discussion, the proposed TS changes do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: Jacob I. Zimmerman, Acting.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

    Date of amendment request: May 31, 2002.
    Description of amendment request: The proposed license amendment 
would revise the Technical Specifications (TSs) to allow the Unit No.1 
core to be operated with a positive moderator temperature coefficient 
(PMTC). TS 3/4.1.1.4, ``Reactivity Control System--Moderator 
Temperature Coefficient (MTC),'' would be changed from the current MTC 
limit of 0x10-\4\ [Delta]k/k/[deg]F to 
+0.2x10-\4\ [Delta]k/k/[deg]F for power levels up to 70 
percent of Rated Thermal Power (RTP) and then ramping lineally from 
+0.2x10-\4\ [Delta]k/k/[deg]F at 70 percent RTP to 
0x10-\4\ [Delta]k/k/[deg]F at 100% RTP. This change is being 
requested to address future core design requirements associated with 
plant operations at higher capacity factors. The amendment would 
include editorial and format changes as well as repagination in order 
to incorporate the revision into the TSs.
    Basis or 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 change from a moderator temperature coefficient 
(MTC) limit of 0 x 10-\4\ [Delta]k/k/[deg]F to a positive 
moderator temperature coefficient (PMTC) of +0.2 x 10-\4\ 
[Delta]k/k/[deg]F does not introduce an initiator of any design 
basis accident or event. The proposed change does not adversely 
affect accident initiators or precursors nor alter the configuration 
of the facility or the manner in which the plant is maintained. 
Thus, the proposed change does not involve a significant increase in 
the probability of an accident previously evaluated.
    The proposed change to a PMTC does 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. The proposed 
change is consistent with the safety analysis assumptions and 
resultant consequences. Accident analyses affected by the proposed 
change have been reanalyzed and all applicable acceptance criteria 
have been met. Thus, the proposed change does not involve a 
significant increase in the 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. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The change to a PMTC does not involve a 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 change to a PMTC does not involve revisions to 
any safety limits or safety system settings that would adversely 
impact plant safety. 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.
    All of the applicable acceptance criteria (i.e., preventing 
reactor coolant system [RCS] or main steam system 
overpressurization, maintaining the minimum departure from nucleate 
boiling ratio [DNBR], preventing core uncovery, preventing fuel 
temperatures from exceeding their limit, preventing clad damage, and 
limiting the number of fuel rods that enter a departure from 
nucleate boiling [DNB] condition) for each of the analyses affected 
by the proposed change continue to be met. The conclusions of the 
Updated Final Safety Analysis Report (UFSAR) remain valid. Thus, 
since the operating parameters and system performance will remain 
within design requirements and safety analysis assumptions, 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.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Sation, Unit 2, Beaver County, Pennsylvania

    Date of amendment request: July 24, 2002.
    Description of amendment request: The proposed license amendment 
would revise the Beaver Valley Power Station, Unit No. 2, (BVPS 2) 
Technical Specifications (TS) Surveillance Requirement (SR) 4.7.1.5 to 
change the valve stroke time limit for full closure of each Main Steam 
Isolation Valve (MSIV) to within 6 seconds from its current 5-second 
limit. The amendment would also replace the quarterly partial stroke 
exercise requirement with criteria to test each MSIV pursuant to 
Specification 4.0.5. TS 4.0.5 requires testing in accordance with 
Section 11 of the American Society of Mechanical Engineering (ASME) 
Code.
    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:


[[Page 58645]]


    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 surveillance criteria for the 
Main Steam Isolation Valves (MSIVs) do not introduce any new 
initiator of a design basis accident. These proposed changes do not 
involve any physical modifications to the MSIVs. 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 proposed frequency change would reduce 
the potential for an (inadvertent) event initiator of full MSIV 
closure and resulting plant transient while retaining a sufficient 
test frequency to identify potential MSIV malfunctions, based on 
industry operating experience. 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. The proposed changes are 
consistent with the safety analyses assumptions and resultant 
consequences. Accident analyses potentially affected by the proposed 
change have been reviewed and all applicable acceptance criteria 
continue to be met. Thus, the proposed changes do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Therefore, the proposed changes do 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?
    No. The proposed change to the surveillance criteria for MSIVs 
do not involve a 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 changes do 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 change to the surveillance criteria for MSIVs 
do not involve revisions to any safety limit or safety system 
settings that would adversely impact plant safety. 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 basis accident conditions within the facility. The 
proposed frequency change would reduce the potential for an 
(inadvertent) event initiator of full MSIV closure and resulting 
plant transient while retaining a sufficient test frequency to 
identify potential MSIV malfunctions, based on industry operating 
experience.
    All of the applicable acceptance criteria for each of the 
analyses affected by the proposed changes continue to be met. The 
conclusions of the Updated Final Safety Analysis Report (UFSAR) 
remain valid. Thus, since the operating parameters and system 
performance will remain within designed requirements and safety 
analysis assumptions, 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.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Sation, Unit 2, Beaver County, Pennsylvania

    Date of amendment request: May 31, 2002.
    Description of amendment request: The proposed license amendment 
would revise the Beaver Valley Power Station (BVPS) Unit No. 2 
Technical Specification (TS) Design Feature 5.3.1, Criticality, where 
the new fuel (fresh fuel) racks enrichment limit specified in Section 
5.3.1.2.a would be increased to 5.00 weight percent (w/o) from its 
current 4.85 w/o limit.
    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 change to the new fuel storage racks enrichment 
limit does not introduce an initiator of any design basis accident. 
The text change on [the] tolerance is added for clarification of the 
criteria associated with [the] new fuel enrichment limit. 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. 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 acceptable limits. The proposed changes are 
consistent with the safety analyses assumptions and resultant 
consequences. Accident analyses potentially affected by the proposed 
change have been reviewed and all applicable acceptance criteria 
continue to be met. Thus, the proposed changes do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Therefore, the proposed changes do 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?
    No. The proposed change to the new fuel storage racks enrichment 
limit and its associated text clarifications do not involve a 
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 changes do 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 change to the new fuel storage racks enrichment 
limit and its associated text clarifications do not involve 
revisions to any safety limit or safety system settings that would 
adversely impact plant safety. 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 basis 
accident conditions within the facility.
    All of the applicable acceptance criteria for each of the 
analyses affected by the proposed changes continue to be met. The 
conclusions of the Updated Final Safety Analysis Report (UFSAR) 
remain valid. Thus, since the operating parameters and system 
performance will remain within design requirements and safety 
analysis assumptions, 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.

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

    Date of amendment request: November 21, 2001, as supplemented 
January 25, 2002, and August 15, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS)

[[Page 58646]]

Surveillance Requirement (SR) 4.0.3 to extend the delay period, before 
entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from ``* * * up to 24 
hours to permit completion of the surveillance when the allowable 
outage time limits of the ACTION requirements are less than 24 hours'' 
to ``* * * up to 24 hours or up to the limit of the specified 
frequency, whichever is greater.'' In addition, the following 
requirement would be added to SR 4.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.'' The proposed amendments are consistent 
with TS Task Force traveler TSTF-358, which has been approved by the 
NRC. The TS Bases will be revised under the licensee's existing TS 
Bases control program to be consistent with TSTF-358. Lastly, a 
proposed administrative change moves two sentences dealing with 
operability requirements from SR 4.0.3 to SR 4.0.1 to make the revised 
TS consistent with the Standard TS for Combustion Engineering plants.
    With regard to the first two changes, the NRC staff issued a notice 
of opportunity for comment in the Federal Register on June 14, 2001 (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the model NSHC 
determination in its application dated November 21, 2001, as 
supplemented January 25, 2002, and August 15, 2002.
    With respect to the administrative changes, the licensee provided 
an additional NSHC determination in its August 15, 2002, supplement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration for the changes associated with 
extending the delay period for a missed surveillance is presented 
below:

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

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

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function. Therefore, this change does 
not involve a significant reduction in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

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

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature and they do 
not affect assumptions contained in plant safety analyses, the 
physical design and/or operation of the plant, nor do they affect 
Technical Specifications that preserve safety analysis assumptions. 
These proposed changes do not change the existing administrative 
controls on performance of Surveillance Requirements. The changes 
only relocate the existing requirements to SR 4.0.1 to closely 
conform to the Standard Technical Specifications. Further, the 
proposed changes do not alter the design, function, or operation of 
any plant component. Therefore, operation of the facility in 
accordance with the proposed amendments would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The changes being proposed are administrative in nature and do 
not introduce a new mode of plant operation or surveillance 
requirement, nor involve a physical modification to the plant. 
Therefore, the design, function, or operation of any plant component 
is not altered. The changes propose to relocate specific controls 
from SR 4.0.3 to SR 4.0.1 to closely conform to the Standard 
Technical Specifications. Therefore, operation of the facility in 
accordance with the proposed amendments would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes conform closely to the industry and NRC 
approved TSTF-358 and relates to the relocation of TS specific 
controls for Surveillance Requirements from SR 4.0.3 to SR 4.0.1. 
The specific controls are not changed only relocated to closely 
conform to the Standard Technical Specifications. Therefore, 
operation of the facility in accordance with the proposed amendments 
would not involve a significant reduction in a margin of safety.


[[Page 58647]]


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

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

    Date of amendment request: August 15, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Section 6.8.4.h, Containment 
Leakage Rate Testing Program, to allow a one-time 5-year extension to 
the current 10-year test interval for the containment integrated leak 
rate test (ILRT). The proposed changes are submitted on a risk-informed 
basis as described in Regulatory Guide 1.174, An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis. The risk-informed analysis 
supporting the proposed changes indicates that the increase in risk 
from extending the ILRT test interval from 10 to 15 years is 
insignificant.

    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) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed amendments of the Technical Specifications add a one 
time extension to the current surveillance interval for Type A 
testing (ILRT). The current test interval of 10 years, based on 
performance history, would be extended on a one time basis to 15 
years from the last Type A test. The proposed extension to Type A 
testing cannot increase the probability of an accident previously 
evaluated since the containment Type A test is not a modification, 
nor a change in the way that plant systems, structures, or 
components (SSC) are operated, and is not an activity that could 
lead to equipment failure or accident initiation. The proposed 
extension of the test interval does not involve a significant 
increase in the consequences of an accident since research 
documented in NUREG-1493, Performance Based Containment Leak-Test 
Program, has found that generically, very few potential leak paths 
are not identified with Type B and C tests. NUREG-1493 concluded 
that an increase in the test interval to 20 years resulted in an 
imperceptible increase in risk. St. Lucie Units 1 and 2 provide a 
high degree of assurance through testing and inspection that the 
containment will not degrade in a manner only detectable by Type A 
testing. Inspections required by the ASME code and the Maintenance 
Rule are performed in order to identify indications of containment 
degradation that could affect leak-tightness. Type B and C testing 
required by 10 CFR part 50 part Appendix J are not affected by this 
proposed extension to the Type A test interval and will continue to 
identify containment penetrations leakage paths that would otherwise 
require a Type A test.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed changes do not result in operation of the facility 
that would create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed 
extension to Type A testing does not create a new or different type 
of accident for St. Lucie because no physical plant changes are made 
and no compensatory measures are being imposed that could 
potentially lead to a failure. There are no operational changes that 
could introduce a new failure mode or create a new or different kind 
of accident. The proposed changes only add a one time extension to 
the current interval for Type A testing and do not change 
implementation aspects of the test.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes would not result in operation of the 
facility involving a significant reduction in a margin of safety. 
The proposed license amendments add a one time extension to the 
current interval for Type A testing. The current test interval of 10 
years, based on historical performance, would be extended on a one 
time basis to 15 years from the last Type A test. The NUREG-1493 
generic study of the effects of extending the Type A test interval 
out to 20 years concluded that there is an imperceptible increase in 
plant risk. Further, the extended test interval would have a minimal 
affect on such risk since Type B and C testing detect over 95 
percent of potential leakage paths. A plant specific risk 
calculation, as part of the CEOG [Combustion Engineering Owners 
Group] joint application report, on this topic obtained results 
consistent with the generic conclusions of NUREG-1493. The overall 
increase in risk contribution was determined as 0.49 percent for 
Unit 1 and 0.30 percent for Unit 2.

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

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: July 23, 2002.
    Description of amendment requests: The proposed amendments would 
revise certain 18 month surveillance requirements by eliminating the 
condition that testing be conducted ``during shutdown,'' or ``during 
the cold shutdown or refueling mode'' (i.e., shutdown conditions).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

Probability of Occurrence of an Accident Previously Evaluated

    The proposed change would eliminate the requirement to perform 
certain 18-month surveillance tests during a shutdown condition. 
These surveillance tests verify that equipment will perform its 
intended safety function of mitigating an accident. Performing the 
surveillance tests during power operation does not affect any 
existing accident initiators or precursors. The proposed change will 
not create any adverse interactions with other systems that could 
result in initiation of a design basis accident. The format and 
capitalization changes are proposed to improve readability and 
appearance, and do not alter any requirements. Therefore, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased.

Consequences of an Accident Previously Evaluated

    The proposed change does not reduce the ability of the 
mitigating equipment to perform its safety function. The [technical 
specification] TS will continue to require the surveillance tests be 
performed on an 18 month periodicity to verify operability. One 
train will be verified as operable prior to testing equipment in the 
other train, thereby making it available to mitigate an accident. 
The accident analyses assume only one train is operable in the event 
of an accident. As a

[[Page 58648]]

result, the ability of the mitigating equipment to perform its 
safety function is unaffected by the proposed change. The format and 
capitalization changes are proposed to improve readability and 
appearance, and do not alter any requirements. Therefore, the safety 
related systems and components that are supported by the equipment 
to mitigate the consequences of an accident are not affected by the 
proposed change.
    In summary, the probability of occurrence and the consequences 
of an accident previously evaluated are not significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not create any new or different 
accident initiators or precursors. The mitigating equipment will 
continue to function as before the change, and will continue to be 
tested at the same surveillance test interval for operability. The 
proposed change does not create any new failure modes for the 
mitigating equipment and does not affect the interaction between the 
equipment and any other system. The format and capitalization 
changes are proposed to improve readability and appearance, and do 
not alter any requirements. Thus, 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 margins of safety applicable to the proposed change are 
those associated with the capability of the mitigating equipment to 
perform its safety function. The proposed change allows the 
surveillance test to be performed during power operation without 
significantly reducing the capability of the mitigating equipment to 
perform in accordance with its safety margin. The format and 
capitalization changes are proposed to improve readability and 
appearance, and do not alter any requirements. Therefore, the 
proposed change does not involve a significant reduction in 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.

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

    Date of amendment requests: July 31, 2002.
    Description of amendment requests: The proposed change will revise 
Diablo Canyon Power Plant (DCPP) Technical Specification (TS) 5.6.6, 
``Reactor Coolant System (RCS) Pressure and Temperature Limits Report 
(PTLR),'' to reference WCAP-14040-NP-A, ``Methodology Used to Develop 
Cold Overpressure Mitigating System Setpoints and RCS Heatup and 
Cooldown Limit Curves,'' as the approved methodology for the PTLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed Technical Specification (TS) changes provide the 
reference for the NRC approved methodology for the Diablo Canyon 
Power Plant (DCPP) Pressure And Temperature Limits Report (PTLR). 
The TS and PTLR were developed using the guidance of NRC Generic 
Letter (GL) 96-03, ``Relocation of the Pressure Temperature Limit 
Curves and Low Temperature Overpressure Protection System Limits,'' 
dated January 31, 1996, which provides guidance on relocating 
reactor coolant system (RCS) pressure/temperature (P/T) limit curves 
and low-temperature overpressure (LTOP) system limits from TS to a 
PTLR. NRC approval of the DCPP specific application of the PTLR 
methodology will allow PG&E [Pacific Gas and Electric Company] to 
use the approved PTLR methodology in the future to calculate new P/T 
and LTOP limits without prior NRC staff approval.
    The proposed PTLR was developed using methodology previously 
approved by the NRC, primarily WCAP-14040-NP-A, Revision 2, 
``Methodology Used to Develop Cold Overpressure Mitigating System 
Setpoints and RCS Heatup and Cooldown Limit Curves,'' dated January 
1996. PG&E has evaluated this methodology and concludes it is 
applicable for use at DCPP. As a result, use of this methodology 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed change completes relocation of the RCS P/T 
and LTOP limits from the TS to the PTLR. The DCPP PTLR submitted 
with this amendment has been developed primarily using the NRC-
approved methodology of WCAP-14040-NP-A, Revision 2.
    The proposed change makes no changes to plant equipment, and 
does not physically alter or change the function of any structures, 
systems or components that could initiate an accident. Through the 
PTLR, it provides operational controls to assure that current RCS P/
T and LTOP limits are not violated. It provides for use of NRC-
approved methodology for changing the RCS P/T and LTOP limits in the 
future without requiring prior NRC approval. As a result, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change completes relocation of the RCS P/T and LTOP 
limits from the TS to the PTLR, and submits the DCPP PTLR 
methodology for NRC approval. The DCPP PTLR submitted with this 
amendment has been developed using the methodology of WCAP-14040-NP-
A, Revision 2, which has previously been approved by the NRC.
    The proposed change makes no changes to plant equipment, and 
does not physically alter or change the function of any structures, 
systems or components that could affect any margin of safety. 
Through the PTLR, it provides operational controls to assure that 
current RCS P/T and LTOP limits are not violated. It provides for 
use of NRC approved methodology for changing the RCS P/T and LTOP 
limits in the future without requiring prior NRC approval. As a 
result, the proposed change has no affect on any margin of safety.
    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: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

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

[[Page 58649]]

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: July 19, 2002.
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification Surveillance Requirement (SR) 
4.0.3 to extend the delay period, before entering a Limiting Condition 
for Operation, following a missed surveillance. The delay period would 
be extended from the current limit of ``* * * up to 24 hours'' to ``* * 
* up to 24 hours or up to the limit of the specified surveillance 
interval, whichever is greater.'' In addition, the following 
requirement would be added to SR 4.0.3: ``A risk evaluation shall be 
performed for any surveillance delayed greater than 24 hours and the 
risk impact shall be managed.'' The proposed amendment would also make 
administrative changes to SRs 4.0.1 and 4.0.3 to be consistent with 
NUREG-1432, Revision 2.
    Date of publication of individual notice in the Federal Register: 
August 22, 2002 (67 FR 54497).
    Expiration date of individual notice: September 23, 2002.

Duke Energy Corporation, et al., (the Licensee) for Operation of the 
Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 50-414 , 
and McGuire Nuclear Station, Units 1 and 2, Docket Nos 50-369 and 50-
370, located in York County, South Carolina and Mecklenburg County, 
North Carolina

    Date of amendment request: October 7, 2001, as supplemented by 
letter dated August 7, 2002.
    Brief description of amendment request: The proposed amendments 
would revise Technical Specification (TS) 5.6.5 regarding the Core 
Operating Limits Report (COLR). TS 5.6.5.a lists the parameters for 
which the limiting values have been relocated by previous TS amendments 
from the TS to the COLR. Specifically, for both Catawba and McGuire 
Nuclear Stations, the amendments would revise the TS 5.6.5.a by (1) 
adding ``60 ppm'' to Item 5.6.5.a.1 regarding the moderator temperature 
coefficient (MTC) surveillance limit for Specification 3.1.3, and (2) 
by adding Item 5.6.5.a.12, ``31 EFPD [effective full-power day] 
surveillance penalty factors for Specifications 3.2.1 and 3.2.2.'' In 
addition, for Catawba Nuclear Station, the amendments would add Item 
5.6.5.a.13, ``Reactor makeup water pumps combined flow rates limit for 
Specifications 3.3.9 and 3.9.2.''
    The limiting values for these parameters were previously relocated 
from the TS to the COLR without the parameter identifier being retained 
in the TS. Inclusion of the parameter identifier in the TS will improve 
consistency between the TS and the COLR. The amendments would also 
change Bases 3.2.1 and 3.2.3 to remove the specific date of the 
referenced topical report.
    Date of publication of individual notice in the Federal Register: 
August 23, 2002 (67 FR 54680).
    Expiration date of individual notice: September 23, 2002.

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

    Date of amendment request: May 14, 2002.
    Brief description of amendment request: The proposed changes would 
modify technical specification (TS) requirements for a missed 
surveillance through revision of Specifications 4.0.1 and 4.0.3. The 
delay period would be extended from the current limit of ``* * * up to 
24 hours to permit the completion of the surveillance when the 
allowable outage time limits of the ACTION requirements are less than 
24 hours'' to ``* * * up to 24 hours or up to the limit of the 
specified Surveillance time interval, whichever is greater.'' In 
addition, the following requirement would be added to Surveillance 
Requirement 4.0.3: ``A risk evaluation shall be performed for any 
Surveillance delayed greater than 24 hours and the risk impact shall be 
managed.'' The proposed revision would also add a TS Bases Control 
Program to the Limerick Generating Station (LGS) TS.
    The proposed amendment would make administrative changes to TS 
6.2.2.g to revise the designation of which manager in the operations 
department shall hold a senior reactor operator license and to TS 
6.5.1.2 to revise the LGS Plant Operations Review Committee (PORC) 
member composition.
    Date of publication of individual notice in the Federal Register: 
August 27, 2002 (67 FR 55041).
    Expiration date of individual notice: September 26, 2002.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

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

    Date of application for amendment: September 10, 2001.
    Brief description of amendment: The amendment revised the 
requirements in Technical Specification Section 3.9, changing the 
number of operable source range monitors (SRMs) from one SRM nearest 
the core alteration to two SRM channels, one with its detector located 
in the core quadrant where core alterations are being performed, and

[[Page 58650]]

another with its detector located in an adjacent quadrant.
    Date of Issuance: September 5, 2002.
    Effective date: As of the date of issuance, to be implemented 
within 30 days of issuance.
    Amendment No.: 229.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in the Federal Register: November 28, 2001 
(66 FR 59501). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated September 5, 2002.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: March 25, 2002.
    Brief Description of amendments: The proposed amendment would 
revise Surveillance Requirement 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance at the Brunswick Steam Electric Plant, Units 1 and 2.
    Date of issuance: August 26, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days from date of issuance.
    Amendment Nos.: 224 and 249.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in the Federal Register: May 28, 2002 (67 FR 
36927). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 1, 2001 as supplemented 
on May 13 and July 1, 2002.
    Brief description of amendment: The amendment modifies the 
Millstone Power Station, Unit No. 3 Technical Specifications to 
increase the emergency diesel generator (EDG) allowed outage time, to 
perform a verification of the offsite circuits within 1 hour prior to, 
or after entering, the condition of either an inoperable offsite source 
or inoperable EDG, to revise the requirements for the pressurizer 
heaters and the pressurizer power operated relief and block valves, and 
to improve the format of the electrical power sources action 
requirements.
    Date of issuance: August 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 210.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in the Federal Register: January 22, 2002 
(67 FR 2920). The May 13 and July 12, 2002, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination and was within the scope of the 
original application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 26, 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: March 26, 2002, as supplemented 
by letter dated June 3, 2002.
    Brief description of amendments: The amendments revise the 
Technical Specification section 1.1, ``Definitions,'' to eliminate 
response time testing requirements for selected sensors and specified 
instrumentation loops for the engineered safety features system and the 
reactor trip system.
    Date of issuance: August 23, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 206/187.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in the Federal Register: April 30, 2002 (67 
FR 21286). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 23, 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: February 20, 2001, as 
supplemented by letters dated July 5, 2001, March 28, 2002, and June 
14, 2002.
    Brief description of amendment: The amendment consists of changes 
to Columbia Generating Station Physical Security Plan pertaining to the 
independent spent fuel storage facility installation (ISFSI).
    Date of issuance: August 27, 2002.
    Effective date: August 27, 2002, and shall be implemented within 30 
days from the date of issuance.
    Amendment No.: 178.
    Facility Operating License No. NPF-21: The amendment revised the 
operating license.
    Date of initial notice in the Federal Register: April 4, 2002 (66 
FR 17966). The July 5, 2001, September 13, 2001, March 28, 2002, and 
June 14, 2002, supplemental letters provided additional clarifying 
information, 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 August 27, 2002.
    The Safety Evaluation contains Safeguards information and is not 
publicly available.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: June 21, 2001, as supplemented 
on February 8, 2002.
    Brief description of amendment: The amendment revises the control 
rod block instrumentation requirements contained in Technical 
Specifications (TS) 2.1.B, Figure 2.1.1, and Tables 3.2.5 and 4.2.5. 
Some of the control rod block trip functions are being relocated to the 
Vermont Yankee Technical Requirements Manual and some of the 
requirements for the retained trip functions are clarified. Two trip 
functions are being added to the TSs.
    Date of issuance: August 27, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 211.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in the Federal Register: March 19, 2002 (67 
FR 12608). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated August 27, 2002.
    No significant hazards consideration comments received: No.

[[Page 58651]]

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi
    Date of application for amendment: June 12, 2002.
    Brief description of amendment: The amendment revises 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 is 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 is 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: August 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 152.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in the Federal Register: July 23, 2002 (67 
FR 48217). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 26, 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: February 15, 2002, as 
supplemented by letter dated June 25, 2002.
    Brief description of amendments: These amendments revised the 
reactor water cleanup system (RWCS) steam leak detection temperature 
isolation actuation instrumentation setpoints contained in Table 3.3.2-
2 concerning items 3.b and 3.c for RWCS area temperature--high and RWCS 
area ventilation differential temperature--high.
    Date of issuance: August 30, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 161/123.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in the Federal Register: June 11, 2002 (67 
FR 40024). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 25, 2002.
    Brief description of amendments: The amendments revise the 
Technical Specifications, Appendix B, ``Environmental Protection Plan 
(Non-Radiological)'' to incorporate by reference the revised terms and 
conditions of the Incidental Take Statement included in the Biological 
Opinion issued by the National Marine Fisheries Service (NMFS) on May 
4, 2001, as modified by NMFS letter dated October 8, 2001. They also 
incorporate administrative revisions necessary to change references to 
the National Pollutant Discharge Elimination System Permit to the 
Wastewater Permit, based on a change in administrative authority over 
these permits.
    Date of issuance: August 28, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 183 and 126.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in the Federal Register: February 19, 2002 
(67 FR 7419). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 28, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 23, 2002, as supplemented 
July 15, 2002.
    Brief description of amendments: Revised Technical Specifications 
to remove the requirement for operability of certain systems when 
handling fuel assemblies that have decayed a sufficient period of time 
such that dose consequences of the postulated fuel handling accident 
remain below the limits of 10 CFR part 100 and the NRC Standard Review 
Plan with these systems unavailable.
    Date of issuance: August 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 184 and 127.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in the Federal Register: June 25, 2002 (67 
FR 42827). The July 15, 2002, supplement did not affect the original 
proposed no significant hazards determination, or expand the scope of 
the request as noticed in the Federal Register. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated August 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 28, 2002.
    Brief description of amendment: The amendment revises Section 6.14, 
``Systems Integrity,'' of the Technical Specifications to eliminate the 
Post Accident Sampling System (PASS) as a potential leakage path 
outside the primary containment. In addition, the amendment supersedes 
the previous requirements for installing and maintaining the PASS, 
which were imposed by NRC confirmatory orders dated March 14, 1983, and 
June 12, 1984.
    Date of issuance: August 26, 2002.
    Effective date: August 26, 2002.
    Amendment No.: 174.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in the Federal Register: July 23, 2002 (67 
FR 48219). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 15, 2002.
    Brief description of amendment: The amendment revised Section 
4.6.4, ``Shock Suppressors (Snubbers),'' following the guidance of 
Generic Letter 90-09, ``Alternative Requirements for Snubber Visual 
Inspection and Corrective Actions,'' dated December 11, 1990.
    Date of issuance: August 28, 2002.
    Effective date: As of the date of its issuance and shall be 
implemented prior to the spring 2003 refueling outage.

[[Page 58652]]

    Amendment No.: 175.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in the Federal Register: April 16, 2002 (67 
FR 18645). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 28, 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: September 27, 2001, as 
supplemented by letter dated May 14, 2002.
    Brief description of amendment: The amendment (1) revises the 
diesel fuel supply volume required for diesel generator (DG) 
operability, (2) clarifies existing wording in the Technical 
Specifications (TS), (3) adds a TS limiting condition for operation 
(LCO), and a TS Surveillance Requirement (SR) regarding the DG air 
receivers, (4) deletes a current TS SR concerning DG starting air 
compressors, and (5) restructures and renumbers the TS LCOs and SR for 
applicability and administrative purposes.
    Date of issuance: August 27, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 129.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in the Federal Register: October 17, 2001 
(66 FR 52801). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 27, 2002.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: May 29, 2002, as supplemented 
July 12, 2002.
    Brief description of amendments: These amendments revise Technical 
Specification 3.8.1, ``AC Sources--Operating,'' to allow portions of 
Surveillance Requirement 3.8.1 to be performed with the units in Mode 
1, 2, 3, or 4. The proposed amendments are consistent with changes made 
to NUREG-1431, Standard Technical Specifications, Westinghouse Plants, 
by Technical Specification Task Force (TSTF) Traveler, TSTF-283, 
Revision 3.
    Date of issuance: August 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 204 and 209.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in the Federal Register: July 9, 2002 (67 FR 
45571). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 29, 2002.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: March 20, 2002.
    Brief description of amendments: The amendments revise Technical 
Specification 3.7.8 to allow the service water (SW) system to be 
operable with five operable SW pumps, provided one unit is in Mode 5 or 
6, or defueled, and the SW system is capable of providing the required 
cooling water flow to required equipment.
    Date of issuance: August 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 205 and 210.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in the Federal Register: May 14, 2002 (67 FR 
34490). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 24, 2001, as supplemented 
April 4, 2002, May 7, 2002, June 17, 2002, July 2, 2002, July 15, 2002, 
and July 25, 2002.
    Brief description of amendment: This amendment increases the spent 
fuel pool storage capacity by replacing all 11 existing rack modules 
with 12 new storage racks. The rerack increases the storage capacity 
from 1,276 storage cells to 1,712 storage cells. The degrading Boraflex 
neutron-absorbing material in the existing racks will be replaced by 
Boral material that will be used in the new racks.
    Date of issuance: August 30, 2002.
    Effective date: August 30, 2002.
    Amendment No.: 160.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in the Federal Register: June 25, 2002 (67 
FR 42810), and repeated on August 20, 2002 (67 FR 53993). The 
supplements listed above contained clarifying information only and did 
not change the initial proposed no significant hazards consideration 
determination or expand the scope of the initial application. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 30, 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: December 10, 2001, as supplemented by 
letter dated May 23, 2002.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) Surveillance Requirements (SRs) 4.0.1 and 4.0.3, 
and incorporate a Bases Control Program in new TS 6.8.3m, in accordance 
with the U. S. Nuclear Regulatory Commission (NRC) staff's position on 
missed surveillances as described in TS Task Force--358, Revision 6. 
The change to SR 4.0.3 extends the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours'' to ``* * * up to 24 hours or up to the limit of the specified 
surveillance interval, whichever is greater.'' In addition, the 
following requirement is added to SR 4.0.3: ``A risk evaluation shall 
be performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.'' In addition to revising SR 4.0.3, part 
of SR 4.0.3 is relocated to SR 4.0.1 and SR 4.0.1 is revised to conform 
to wording contained in the improved Standard TSs.
    Date of issuance: August 27, 2002.
    Effective date: August 27, 2002.
    Amendment Nos.: Unit 1-141; Unit 2-129.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in the Federal Register: February 5, 2002 
(67 FR 5337). The May 23, 2002, supplemental letter provided clarifying 
information that was within the scope of the original

[[Page 58653]]

Federal Register notice and did not change the staff's initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated August 27, 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: November 29, 2001, as 
supplemented June 18, 2002.
    Brief description of amendment: These amendments establish a new 
operating domain for the containment partial pressure.
    Date of issuance: September 5, 2002.
    Effective date: As of the date of issuance and shall be implemented 
at the end of the Cycle 16/17 refueling outage for Unit 1, and at the 
end of the Cycle 15/16 refueling outage for Unit 2.
    Amendment Nos.: 232/214.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications.
    Date of initial notice in the Federal Register: April 30, 2002 (67 
FR 21295). The June 18, 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 amendments is 
contained in a Safety Evaluation dated September 5, 2002.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit 2, Louisa County, Virginia

    Date of application for amendment: February 11, 2002, as 
supplemented May 16, 2002.
    Brief description of amendment: This amendment revises the Facility 
Operating License (FOL) to allow the operation of one lead test 
assembly containing zirconium-based alloy for one cycle, with a lead 
rod burnup not to exceed 75,000 MWD/MTU.
    Date of issuance: September 4, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 213.
    Facility Operating License No. NPF-7: Amendment changes the FOL.
    Date of initial notice in the Federal Register: April 30, 2002 (67 
FR 21296). The May 16, 2002, supplement contained clarifying 
information only, and did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the initial application. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated September 4, 2002.
    No significant hazards consideration comments received: No

    Dated at Rockville, Maryland, this 6th day of September, 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-23358 Filed 9-16-02; 8:45 am]
BILLING CODE 7590-01-P