[Federal Register Volume 66, Number 229 (Wednesday, November 28, 2001)]
[Notices]
[Pages 59498-59518]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-29446]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

    Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.)

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 5 through November 16, 2001. The 
last biweekly notice was published on November 14, 2001 (66 FR 57116).

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

[[Page 59499]]

within 30 days after the date of publication of this notice will be 
considered in making any final determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By December 28, 2001, 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, which is 
available at the NRC's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically 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/NRC/ADAMS/index.html. 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.
    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 Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for

[[Page 59500]]

public inspection at the Commission's Public Document Room, located at 
One White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Assess and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or 
if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document room (PDR) Reference staff at 1-800-
397-4209, 304-415-4737 or by email to [email protected].

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

    Date of amendment request: April 4, 2001, as supplemented on 
October 12, 2001.
    Description of amendment request: The proposed amendment request 
would delete Technical Specifications (TSs) 5.3.1.B and 5.3.1.C. These 
TSs restrict the handling of heavy loads over irradiated fuel stored in 
the storage pool. The basis for deleting these TSs is the upgrade of 
the reactor building crane and associated handling systems to a single-
failure proof system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    The proposed amendment does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Until August 2000, the reactor building crane was not single-
failure-proof. For heavy load handling associated with the spent fuel 
pool, Oyster Creek was consistent with Section 5.1.4(2) of NUREG-0612: 
``The effects of heavy load drops in the reactor building should be 
analyzed to show that the evaluation criteria of Section 5.1 are 
satisfied.'' An alternative to this is Section 5.1.4(1): ``The reactor 
building crane, and associated lifting devices used for handling of 
heavy loads, should satisfy the single-failure-proof guidelines of 
Section 5.1.6 of this report.'' The upgraded crane and handling systems 
satisfy the guidelines of Section 5.1.6. Therefore, the licensing basis 
for the reactor building crane with regard to its use in handling heavy 
loads above the spent fuel storage pool is being revised to include 
Section 5.1.4(1) of NUREG-0612 in addition to 5.1.4(2).
    The cask drop protection system was required with the original 
crane because the load drop analysis will yield unacceptable 
consequences to the spent fuel storage pool (SFSP) structure. The cask 
drop protection system (CDPS) serves to mitigate the consequences of a 
cask drop accident involving the original crane which complied with 
NUREG-0612 Phase I. The upgraded single-failure-proof crane satisfies 
the criteria of NUREG-0612 Section 5.1.6. Therefore, the reactor 
building crane eliminates reliance on the design function of the CDPS 
because the probability of a heavy load drop is very low.
    With the proposed revisions to the TSs, the evaluation criteria of 
NUREG-0612, Section 5.1 is met with a single-failure-proof crane that 
satisfies the guidelines of Section 5.1.6 or with consequence analyses 
that satisfies Section 5.1.4(2).
    The proposed TS revisions do not significantly change the potential 
for unacceptable consequences to the plant in conducting heavy load 
handling above the SFSP because the probability of a load drop accident 
caused by use of the reactor building crane has been reduced to where 
it is very unlikely, and therefore, can be considered not credible 
within regulatory accepted standards.
    Therefore, the proposed TS revisions do 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 CDPS was installed in the Oyster Creek SFSP to mitigate the 
effects of a cask drop when the reactor building crane was not single-
failure proof. The CDPS acts as a hydraulic dashpot to limit the 
velocity of a falling cask to attenuate impact forces to within 
acceptable levels. The CDPS structure cannot be removed from the spent 
fuel pool without eliminating its functional requirement. The use of 
the CDPS increases the duration of cask lifts and exposure to 
personnel. Therefore, eliminating the complications caused by the use 
of the CDPS together while improving the reliability of the crane and 
associated systems does not create the possibility of a new or 
different kind of accident.
    Therefore, operation of the facility in accordance with the 
proposed license amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed TS change will remove the load limit over the SFSP and 
CDPS restrictions when the reactor building crane is used with single-
failure-proof handling systems that comply with criteria in Section 
5.1.6 of NUREG-0612.
    The reactor building crane was upgraded to single-failure-proof in 
compliance with NUREG-0554. The upgraded crane and handling system is 
in compliance with NUREG-0612, Sections 5.1.1 and 5.1.6. The NRC in 
NUREG-0612, Section 5.2 documented their review of the potential 
consequences of a load drop when handled by a single-failure-proof 
crane using single-failure-proof rigging compared with other 
alternatives and concluded as follows:
    ``The likelihood for unacceptable consequences in terms of 
excessive releases of gap activity or potential for criticality due to 
accidental dropping of postulated heavy loads after Receptionist (OWFN 
and TWFN) implementation of the guidelines of Section 5.1 is very 
low.''
    Therefore, there is a very minimal chance of a load drop that could 
result in consequences that exceed the regulatory accepted standards 
when the load is handled by a single-failure-proof crane and handling 
system, and performed in accordance with Section 5.1 of NUREG-0612. A 
single-failure-proof crane design incorporates the applicable design 
basis event that in this case is a seismic event. A load drop is of 
such low probability that it is considered unlikely when it is handled 
with the reactor building crane because the crane and its handling 
systems satisfy the NUREG-0612 criteria for a single-failure-proof 
crane. Therefore, any load lifted over the SFSP using the reactor 
building crane, and adhering to NUREG-612 Phase I guidelines has a very 
low probability of falling into the spent fuel pool accidentally or as 
a result of a design basis event.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: L. Raghavan, Acting.

[[Page 59501]]

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

    Date of amendment request: September 10, 2001.
    Description of amendment request: The proposed change would revise 
the requirement for the source range monitor (SRM) operability during 
core operations. The proposed change would require two SRM channels to 
be operable, one with its detector located in the core quadrant where 
core alterations are being performed, and another with its detector 
located in an adjacent quadrant.
    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 TS [Technical Specification] change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed change revises Technical Specification 3.9.D for 
source range monitor operability requirements during core 
alterations. The only accident described in the Final Safety 
Analysis Report (FSAR) while the plant is in Cold Shutdown or 
Refueling is a fuel handling (dropped bundle) accident. The proposed 
change involves equipment that is not involved in the mitigation or 
prevention of a fuel handling accident as described in FSAR. 
Therefore, the change to SRM operability requirements does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to TS 3.9.D does not involve any physical 
alteration of plant equipment or system configuration. Core 
reactivity and reactivity control functions are not affected, and 
adequate reactivity monitoring capability is maintained. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed change to TS 3.9.D affects the operability 
requirements for source range monitors during core alterations. The 
SRMs do not perform any required functions for mitigating the 
consequences of an accident. The current specification only requires 
one operable SRM. The proposed specification will ensure redundant 
monitoring is available to detect changes in the reactivity 
condition of the core by requiring the operability of at least two 
source range monitors. This will provide adequate capability for 
detecting an inadvertent criticality. 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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: September 11, 2001
    Description of amendment request: The purpose of the proposed 
revision to the Technical Specifications (TSs) is to delete the cycle-
specific footnote for the Safety Limit Minimum Power Critical Ratio 
(SLMCPR).
    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 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The derivation of the cycle specific SLMCPR limit for 
incorporation into the Technical Specification, and its use to 
determine cycle specific thermal limits, has been performed using 
the methodology discussed in ``General Electric Standard Application 
for Reactor Fuel,'' NEDE-24011-P-A-13, and Amendment 25. Amendment 
25 was approved by the NRC in a Safety Evaluation Report dated March 
11, 1999. The footnote to Technical Specification 2.1.A is being 
deleted. The footnote associated with the Technical Specification 
2.1.A was originally included to ensure that the SLMCPR was only 
applicable for the identified cycle because Amendment 25 was not yet 
NRC approved. Amendment 25 has subsequently been approved. 
Therefore, this footnote is no longer necessary. The footnote was 
for information only, and has no impact on the design or operation 
of the plant. Cycle-specific SLMCPR values will continue to be 
developed in accordance with NRC approved methods, which ensures 
that applicable regulatory requirements are met [. . .]
    Therefore, this change does 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 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change deletes the footnote contained in Technical 
Specification 2.1.A as the result of the NRC approval of Amendment 
25 to NEDE-24011-P-A. This change does not affect the design or 
operation of any plant structures, systems or components. Cycle-
specific SLMCPR values will continue to be developed in accordance 
with NRC approved methods, which ensures that applicable regulatory 
requirements are met. Changes to the SLMCPR value specified in the 
Technical Specification will require prior NRC approval [. . .]
    Therefore, this change does 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 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed change deletes the footnote contained in Technical 
Specification 2.1.A as the result of the NRC approval of Amendment 
25 to NEDE-24011-P-A. Cycle-specific SLMCPR values will continue to 
be developed in accordance with NRC approved methods as specified in 
the Technical Specifications. These methods ensure that applicable 
regulatory requirements are met. Changes to the SLMCPR value 
specified in the Technical Specifications will require prior NRC 
approval [. . .]
    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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendments request: September 11, 2001.
    Description of amendments request: The amendments would allow the 
non-operating shutdown cooling loop to be declared inoperable for a 
period up to 2 hours for surveillance testing in MODE 6. The request is 
based on Technical Specification Task Force Traveler Number 361, 
Revision 2.

[[Page 59502]]

    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 amendment would add a note to the limiting 
condition of operation (LCO) of Technical Specification 3.9.5, 
Shutdown Cooling (SDC) and Coolant Circulation--Low Water Level, 
that would permit one required SDC loop to be declared inoperable 
for a period of up to 2 hours for surveillance testing, provided the 
other SDC loop is OPERABLE and in operation.
    Allowing the non-operating SDC loop to be declared inoperable in 
accordance with the proposed amendment does not involve a 
significant increase in the probability of an accident previously 
evaluated because the SDC system is not an accident initiator of any 
previously evaluated accidents. Because the SDC system does not 
initiate any previously analyzed accidents, it cannot increase the 
probability of these accidents occurring.
    Furthermore, allowing the non-operating SDC loop to be declared 
inoperable in accordance with the proposed amendment does not 
involve a significant increase in the consequences of an accident 
previously analyzed because only one operating SDC loop is necessary 
to perform the SDC system function of removing decay heat from the 
reactor core.
    The proposed amendment does not represent a change to the design 
of the facility. Nor does the proposed amendment prevent the safety 
function of the shutdown cooling system from being performed. The 
proposed amendment does not alter, degrade, or prevent actions 
described or assumed in any accident described in the PVNGS Updated 
Final Safety Analysis Report (UFSAR) from being performed. 
Therefore, since the SDC system is not an accident initiator and 
because only one SDC loop is necessary to perform the design 
function, the proposed amendment would not significantly increase 
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 amendment would add a note to the limiting 
condition of operation (LCO) of Technical Specification 3.9.5, 
Shutdown Cooling (SDC) and Coolant Circulation--Low Water Level, 
that would permit one required SDC loop to be declared inoperable 
for a period of up to 2 hours for surveillance testing, provided the 
other SDC loop is OPERABLE and in operation. Allowing the non-
operating SDC loop to be declared inoperable in accordance with the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because: (1) The proposed amendment does not represent a change to 
the design of the plant, (2) the proposed amendment does not involve 
the installation of new or different equipment, (3) the proposed 
amendment does not alter the methods for operating plant equipment, 
and (4) the proposed amendment does not affect any other safety 
related equipment. Therefore, the proposed amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed amendment would add a note to the limiting 
condition of operation (LCO) of Technical Specification (TS) 3.9.5, 
Shutdown Cooling (SDC) and Coolant Circulation--Low Water Level, 
that would permit the non-operating SDC loop to be declared 
inoperable for a period of up to 2 hours for surveillance testing in 
MODE 6, when the water level is less than 23 feet above the top of 
the reactor vessel flange, provided the other SDC loop is OPERABLE 
and in operation. Allowing the non-operating SDC loop to be declared 
inoperable in accordance with the proposed amendment does not 
involve a significant reduction in a margin of safety because the 
operating SDC loop provides sufficient decay heat removal capacity. 
The proposed change does not impact the operating SDC loop. In the 
unlikely event that the operating SDC loop becomes inoperable 
concurrent with the inoperability of the non-operating SDC loop 
allowed by the proposed note, adequate controls exist within the TS 
3.9.5 Required Actions to ensure adequate decay heat removal. In 
addition, if the operating SDC loop fails, operator action to 
restore the SDC loop being tested to OPERABLE status and place that 
SDC loop in operation will be timely such that adequate decay heat 
removal capability is maintained. Therefore, the proposed amendment 
does not involve a significant reduction in a margin of safety.
    Based on the responses to these three criteria, Arizona Public 
Service Company (APS) has concluded that the proposed amendment 
involves no significant hazard consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendments request: November 7, 2001.
    Description of amendments request: The proposed license amendments 
would revise Technical Specification (TS) 3.1.4, ``Control Rod Scram 
Times'' and TS 5.5.10, ``Technical Specifications Bases Control 
Program.'' TS 3.1.4 would be revised to better delineate the 
requirements for testing control rod scram times following refueling 
outages. TS 5.1.10 would be revised to reference Title 10 of the Code 
of Federal Regulations (10 CFR) Section 50.59. This license amendment 
application incorporates the NRC-approved Technical Specification Task 
Force (TSTF) Item 222, Revision 1, ``Control Rod Scram Time Testing,'' 
and TSTF Item 364, Revision 0, ``Revision to TS Bases Control Program 
to Incorporate Changes to 10 CFR 50.59.''
    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 license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to adopt TSTF-222, Revision 1, is an 
administrative clarification of existing Technical Specification 
requirements regarding scram time testing requirements for control 
rods. The current wording of Surveillance Requirement 3.1.4.1 
requires each control rod to be tested if any fuel movement occurs 
in the reactor pressure vessel. Surveillance Requirements 3.1.4.3 
and 3.1.4.4 require only the affected control rods to be tested. The 
NRC-approved TSTF-222, Revision 1, clarifies that post-refueling 
scram time testing of control rods only applies to control rods 
affected by work activities. The requirement to test all control 
rods following routine refueling outages remains unchanged. As such, 
there is no effect on initiators of analyzed events or assumed 
mitigation of accidents or transients. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to adopt TSTF-364, Revision 0, is an 
administrative change to provide consistency between the Technical 
Specification requirements for the Technical Specification Bases 
Control Program and the regulatory requirements of Title 10, Section 
50.59 of the Code of Federal Regulations, as revised by the NRC on 
October 4, 1999. The change will have no affect on the initiators of 
analyzed events or assumed mitigation of accidents or transients.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 59503]]

    The proposed changes to adopt TSTF-222, Revision 1 and TSTF-364, 
Revision 0, do not involve a physical alteration of the plant, add 
any new equipment, or require any existing equipment to be operated 
in a manner different from the present design. Therefore, the 
proposed changes do not create a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed change to adopt TSTF-222, Revision 1, will not 
reduce a margin of safety because it has no effect on any safety 
analysis assumptions. The proposed license amendment implements an 
administrative clarification to better delineate the requirements 
for scram time testing control rods following refueling outages and 
for control rods requiring testing due to work activities. The 
requirement to test all control rods following a routine refueling 
outage remains unchanged. As such, the proposed change does not 
involve a significant reduction in the margin of safety.
    The proposed change to adopt TSTF-364, Revision 0, is an 
administrative change to provide consistency between the Technical 
Specification requirements for the Technical Specification Bases 
Control Program and the regulatory requirements of Title 10, Section 
50.59 of the Code of Federal Regulations, as revised by the NRC on 
October 4, 1999. The change will not reduce the margin of safety 
because the change has no effect on any safety analyses assumptions. 
Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: August 24, 2001.
    Description of amendment request: The proposed amendment would 
delete the Technical Specification (TS)-required action which, in the 
event of inoperability of the oscillation power range monitor (OPRM) 
trip function, limits plant operation above 25-percent power to 120 
days. Instead, continued plant operation would be allowed if a TS-
required action is taken to implement an alternate method to detect and 
suppress thermal-hydraulic instability oscillations.
    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 OPRM function is not considered as an initiator of any 
previously analyzed accident. Therefore, this proposed change does 
not significantly increase the probability of such accidents. This 
proposed change would allow the use of existing well-established 
alternate methods to detect and suppress the thermal hydraulic 
instability oscillations. Considering that multiple Boiling Water 
Reactor plants, including Fermi 2, have satisfactorily operated 
using alternate stability monitoring methods for extended periods of 
time prior to the installation of OPRM systems, it is concluded that 
these measures are adequate. Therefore, the consequences of a 
previously analyzed accident would not be significantly increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    This proposed change would allow the use of an existing 
alternate method to detect and suppress thermal hydraulic 
instability oscillations to continue to operate the reactor above 
25% power in the event of the inoperability of the OPRM system. 
Considering that multiple Boiling Water Reactor plants, including 
Fermi 2, have satisfactorily operated using alternate stability 
monitoring methods for extended periods of time, it is concluded 
that these measures are adequate, and that the proposed change does 
not significantly reduce the margin of safety.

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

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: October 23, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) surveillance requirement (SR) 
3.8.4.1 to change limits for the battery terminal voltage when on a 
float charge for 125 VDC station battery 31 following the replacement 
of this battery in early 2002. The proposed amendment would also revise 
the applicable TS Bases section.
    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 License Amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed TS SR change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The newly installed battery 31 will consist of 
59 cells, instead of the presently installed 58-cell battery. An 
additional cell will be added to 31 Battery in order to provide an 
acceptable design margin for future load addition to this battery.
    The resulting change in the minimum 31 Battery terminal voltage 
on float charge to 125.7 V is due to the additional cell added. This 
new value will ensure that the 31 Battery is properly verified to be 
functional to meet its design requirements. Calculations 
demonstrated in IP3-ECCF-845 indicate that 31 Battery DC circuit 
coordination is not affected by the proposed replacement of the 
existing battery with a 59-cell battery. The proposed TS SR change 
does not affect accident initiators or precursors, nor do they alter 
design assumptions for the systems or components used to mitigate 
the consequences of an accident as analyzed in Chapter 14 of the IP3 
UFSAR [Indian Point 3 Updated Final Safety Analysis Report].
    2. Does the proposed License Amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. This TS SR change for 31 Battery is based upon replacement 
of the 31 Battery with a new 59-cell battery. This new battery 31 is 
at least equivalent to the existing 58-cell 31 Battery. This new 31 
battery, with the added cell, provides an acceptable design margin 
to the 31 Battery. Battery 31 circuit coordination is not adversely 
affected by the addition of this new battery with 59 cells. The 
proposed changes to this TS SR do not introduce any new accident 
initiators or precursors, or any new design assumptions for those 
components used to mitigate the consequences of an accident.

[[Page 59504]]

    3. Does the proposed License Amendment involve a significant 
reduction in a margin of safety?
    No. During the replacement of the existing 31 battery with a new 
59-cell battery and the subsequent TS SR change that verifies higher 
minimum terminal voltage on float charge, the new 31 battery and the 
requirements associated with verifying its design functionality will 
not involve a significant reduction in the margin of safety. The 
replacement 31 Battery is at least equivalent to the existing 
battery. The additional cell in the proposed new 59-cell battery 
provides an acceptable design margin, which will be 120% for 31 
battery with 59 cells. The increase in the number of cells from 58 
to 59 will result in a higher 31 Battery terminal voltage on float 
charge. This proposed TS SR simply documents the verification of 
this new minimum voltage value. The minimum terminal voltage value 
for the new 32 Battery will not change nor be impacted by this TS 
change. Accordingly, there is no 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. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: L. Raghavan (Acting).

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

    Date of amendment request: October 15, 2001.
    Description of amendment request: The proposed change to Technical 
Specification 3.4.7 limits Reactor Coolant System activity permitted by 
the ACTION statement to 60 microcuries per gram (Ci/gm) at all 
power levels. The letdown line break accident analysis in the Final 
Safety Analysis Report is also changed to reflect revised dose 
consequences.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response:
    The proposed change to the Technical Specifications (TS) 
conservatively limits Reactor Coolant System (RCS) activity 
permitted by Action Statement 3.4.7.a to 60 Ci/gm at all 
reactor power levels. The proposed change to the Final Safety 
Analysis Report (FSAR) Section 15.6.3.1 revises the letdown line 
break accident analyses.
    The probability of a previously evaluated accident is not 
affected by this change because the pre-existing iodine spike is not 
an accident initiator and the FSAR change does not affect any plant 
Structure, Systems, or Component (SSC) but merely determines the 
consequences of the previously evaluated accident.
    This TS change is conservative in that it will reduce the 
accident consequences for events occurring at lower power levels.
    The proposed FSAR change meets the original SER [Safety 
Evaluation Report] acceptance criteria with the exception of the 
Exclusion Area Boundary (EAB) accident induced iodine spiking 
thyroid dose. The SRP [Standard Review Plan] acceptance criteria for 
the EAB accident induced iodine spiking thyroid dose is a small 
fraction of the 10 CFR [Part] 100 limits (30 rem). The proposed 
change falls well within 10 CFR [Part] 100 limits (75 rem).
    The EAB accident induced iodine spiking thyroid dose 
consequences are considered acceptable and reasonable for the 
following reasons:
     The letdown line break event starting from the most 
limiting parameters allowed by the TS LCO [Limiting Conditions for 
Operation] on RCS activity, pressure, temperature, primary to 
secondary leakage, and proceeding unmitigated for 30 minutes is 
highly unlikely. The additional use of conservative assumptions such 
as an iodine spiking factor of 500, maximum bounding letdown flow, 
worst case 95 percentile atmospheric dispersion factors, flashing 
fraction based on 560  deg.F even though the break flow would travel 
through the regenerative heat exchanger and cool down, no activity 
plate out, no ground deposition, and no activity decay in the 
transit to the exclusion area boundary significantly increases the 
overall conservative nature of the calculation.
     Currently, FSAR Table 15.6-4 lists the 'Realistic' EAB 
thyroid dose as 0.46 rem. The realistic dose is based upon no iodine 
spike, 50 percentile X/Q [atmospheric dispersion factor], and 0.12% 
failed fuel RCS activity. The best estimate dose consequences using 
the new analysis methodology with the normal plant operating 
parameters would remain below 0.46 rem even for the accident induced 
iodine spiking event.
     The new analysis accident induced iodine spiking 
results would remain below the SRP acceptance criteria if any one of 
the following normal plant operating parameters were used: RCS 
steady state activity, iodine spiking factor, letdown flow, or 
atmospheric dispersion factors.
    The letdown line break consequences are considered acceptable 
due to the unlikeliness of the event and conservative nature of the 
analyses. The `no iodine spike' results remain within a small 
fraction of the 10 CFR [Part] 100 limits; the `accident induced 
iodine spike' results fall well within the 10 CFR [Part] 100 limits; 
and the `pre-existing iodine spike' results are within the 10 CFR 
[Part] 100 limits.
    Therefore, this change does not involve a significant increase 
in the probability or consequence of any accident previously 
evaluated.
    2. Will the operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response:
    The probability of a new or different accident is not affected 
by this change because the pre-existing iodine spike is not an 
accident initiator and the FSAR change does not affect any plant 
Structure, Systems, or Components (SSC) but merely determines the 
consequences of the previously evaluated accident.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will the operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response:
    The TS change is more limiting in that it will reduce the 
accident consequences for events occurring at lower plant levels.
    The proposed FSAR change meets the original SRP acceptance 
criteria with the exception of the Exclusion Area Boundary (EAB) 
accident induced iodine spiking thyroid dose. The SRP acceptance 
criteria for the EAB accident induced iodine spiking thyroid dose is 
a small fraction of the 10 CFR [Part] 100 limits (30 rem). The 
proposed change falls well within 10 CFR [Part] 100 limits (75 rem).
    The EAB accident induced iodine spiking thyroid dose 
consequences are considered not to be a significant reduction in the 
margin of safety for the following reasons.
     The letdown line break event starting from the TS LCO 
on RCS activity, pressure, temperature, primary to secondary 
leakage, and proceeding unmitigated for 30 minutes is highly 
unlikely. The additional use of conservative assumptions such as an 
iodine spiking factor of 500, maximum bounding letdown flow, worst 
case 95 percentile atmospheric dispersion factors, flashing fraction 
based on 560  deg.F even though the break flow would travel through 
the regenerative heat exchanger and cool down, no activity plate 
out, no ground deposition, and no activity decay in the transit to 
the exclusion area boundary significantly increases the overall 
conservative nature of the calculation.
     The FSAR Table 15.6-4 lists the ``Realistic'' EAB 
thyroid dose as 0.46 rem. The realistic dose is based upon no iodine 
spike, 50 percentile X/Q, and 0.12% failed fuel RCS activity. The 
best estimate dose consequences using the new analysis methodology 
with the normal plant operating parameters would remain below 0.46 
rem even for the accident induced iodine spiking event.
     The new analysis accident induced iodine spiking 
results would remain below the SRP acceptance criteria if any one of 
the following normal plant operating parameters were used: RCS 
steady state activity, iodine spiking factor, letdown flow, or 
atmospheric dispersion factors.

[[Page 59505]]

    The letdown line break consequences are considered acceptable 
due to the unlikeliness of the event and conservative nature of the 
analyses. The ``no iodine spike'' results remain within a small 
fraction of the 10 CFR [Part] 100 limits; the ``accident induced 
iodine spike'' results fall well within the 10 CFR [Part] 100 
limits; and the ``pre-existing iodine spike'' results are within the 
10 CFR [Part] 100 limits.

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

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

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

    Date of amendment request: September 21, 2001.
    Description of amendment request: The proposed amendment would 
revise the Reactor Core Safety Limit (SL) for peak fuel centerline 
temperature from less than or equal to 4700  deg.F (i.e., the current 
TS limit) to the design basis fuel centerline melt temperature of less 
than 5080  deg.F, for unirradiated fuel, decreasing by 58  deg.F per 
10,000 Megawatt-Days per Metric Tonne Uranium (MWD/MTU) burnup.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The use of high burnup rods or assemblies will not increase the 
probability of any accident previously evaluated. These high burnup 
rods or assemblies will continue to satisfy all fuel mechanical, 
nuclear, thermal-hydraulic, and transient analysis design criteria.
    Fuel type is not directly related to the probability of any 
previously evaluated accidents; however, adhering to applicable 
design criteria and standards precludes challenges to components and 
systems that could increase the probability of an accident. The high 
burnup fuel rods will continue to satisfy the Specified Acceptable 
Fuel Design Limits (SAFDLs) specified in the Westinghouse Topical 
Report, WCAP-12488-A, ``Westinghouse Fuel Criteria Evaluation 
Process,'' which was approved by the Nuclear Regulatory Commission 
(NRC) on July 27, 1994. The clad integrity of the four high burnup 
rods in the LTA will be maintained as the LTAs will be placed in 
non-limiting core locations as permitted by TS 4.2.1 and will 
continue to meet the safety parameter requirements. In addition, the 
acceptability of using the four high burnup rods in an LTA is 
evaluated in the Byron Station, Unit 2, Cycle 10 Reload Safety 
Evaluation which is supported by Westinghouse Topical Report, 
``Extended Burnup Operation Assessment for the VANTAGE+ Design in 
Byron, Unit 2, Cycle 10,'' dated March 2001.
    It has been shown in Westinghouse Topical Report, WCAP-12610-P-
A, ``VANTAGE+ Fuel Assembly Reference Core Report,'' approved by the 
NRC in April 1995, that even though there are variations in core 
inventories of isotopes due to extended burnup up to 75,000 MWD/MTU, 
there are no significant increases of isotopes that are major 
contributors to accident doses. It is worthy to note that, at higher 
burnups, there is a reduction in certain isotopes that are major 
dose contributors under accident situations (e.g., Kr-88). With only 
four high burnup rods in the entire core, any variation of isotopes 
will be extremely small. Thus, the radiation dose limitations of 10 
CFR [Part] 100, ``Reactor Site Criteria,'' will not be exceeded.
    The bases for establishing the fuel centerline melt temperature 
are discussed in WCAP-12610-P-A, noted above, and implemented by 
Westinghouse Topical Report WCAP-14483-A, ``Generic Methodology for 
Expanded Core Operating Limits Report,'' approved by the NRC on 
January 19, 1999. These methodologies and associated analyses 
confirm that the present analytical limits for all accidents will be 
maintained.
    Based on this evaluation, it is concluded that the proposed TS 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Do the proposed TS changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    As required by WCAP-12488-A, the LTA with the four high burnup 
rods must satisfy the five guidelines accepted by the NRC. These 
guidelines are as follows:
     Design of LTAs are mechanically and hydraulically 
compatible with existing fuel
     Peaking factors meet the TS limits
     NRC approved/accepted safety/design methods and codes 
are used
     No SAFDLs are exceeded
     Not more than eight LTAs per core are inserted
    As previously noted, TS 4.2.1 allows the use of a limited number 
of LTAs in nonlimiting core regions.
    The use of high burnup rods or assemblies will comply with WCAP-
12488-A and TSs. All safety evaluations in support of using high 
burnup rods or assemblies have been performed in accordance with 
accepted methodologies.
    In support of proposed High Burnup LTA Programs in the industry, 
the NRC has requested fuel characterization inspections prior to 
high burnup irradiation. LTA M09E, (i.e., the assembly containing 
the high burnup fuel rods at Byron Station) was subjected to fuel 
characterization inspections prior to operation in Byron Station, 
Unit 2, Cycle 10. These inspections included assembly growth, rod 
growth, assembly bow, peripheral rod oxidation, grid growth, grid 
oxidation, guide thimble inner diameter oxidation, grid cell size, 
crud scraping, single rod exams for the high burnup rods, 
profilometry, and pellet-to-pellet gap measurements using a Gamma 
Scanner instrument. All parameters inspected were found to be 
acceptable.
    By performing the above inspection regimen, the demonstrated 
adherence to the inspection standards and acceptance criteria 
precludes the potential for new risks to components and systems that 
could introduce a new type of accident.
    Based on this evaluation, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    There is no significant reduction in the margin of safety due to 
the proposed change. The current TS Safety Limit (SL) 2.1.1.3 states 
that ``In MODES 1 and 2, the peak fuel centerline temperature shall 
be maintained  4700  deg.F.'' The TS Safety Limit Bases 
states that overheating of the fuel is prevented by maintaining the 
steady state peak Linear Heat Rate (LHR) below the level at which 
fuel centerline melting occurs. Fuel centerline melting occurs when 
the local LHR, or power peaking, in a region of fuel is high enough 
to cause the fuel centerline temperature to reach the fuel melting 
point.
    WCAP-14483-A conservatively states that the fuel centerline 
temperature limit has been established based on the melting 
temperature for Uranium Dioxide (UO2) fuel of 5080 
deg.F, decreasing by 58  deg.F per 10,000 MWD/MTU of burnup. Based 
on the WCAP-14483-A equation, a burnup of approximately 65,500 MWD/
MTU could be accrued before the melting temperature would 
academically reach the current TS SL of 4700  deg.F.
    Westinghouse has evaluated the fuel centerline temperatures for 
the Byron Station and Braidwood Station reactor cores under uprated 
power conditions. This evaluation shows that the high burnup rods' 
temperatures would remain below both the current SL of 4700  deg.F 
and the proposed WCAP-14483-A equation (i.e., the proposed SL) for 
fuel melting temperatures under extended burnup conditions past 
75,000 MWD/MTU. Thus, fuel melting will not occur in the LTA high 
burnup rods.
    The insertion of the four high burnup rods does not impact any 
other TS. The LTA has been designed to operate within the SAFDLs and 
will therefore have sufficient safety margins. Furthermore, the high 
burnup LTA will satisfy the five guidelines specified in WCAP-12488-
A approved by the NRC. The high burnup LTA will comply with TS 4.2.1 
by being placed in a nonlimiting core region.
    Based on the above discussion, changing the fuel centerline melt 
temperature from the

[[Page 59506]]

existing 4700  deg.F to an equation consistent with the design basis 
for fuel melt temperature will not significantly reduce the margin 
of safety. The analysis shown in WCAP-12610-P-A indicates that the 
minimum margin to safety occurs at fuel assembly Beginning of Life 
(BOL). The evaluation in WCAP-12610-P-A demonstrates that margin of 
safety with respect to the proposed SL equation remains sufficient 
for fuel burnups up to 75,000 MWD/MTU.
    Based on this evaluation, the proposed TS changes do not involve 
a significant reduction in a margin of safety.
    Conclusion: Based upon the above analyses and evaluations, we 
have concluded that the proposed change to the TS involves no 
significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve 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.

Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of amendment request: September 14, 2001.
    Description of amendment request: Exelon proposed to extend the use 
of the pressure temperature limits specified in Technical Specification 
(TS) Figure 3.4.6.1-1, ``Minimum Reactor Vessel Metal Temperature vs. 
Reactor Vessel Pressure,'' through Cycle 10 of operation, currently 
scheduled to end April 2004. Exelon also proposed to modify TS Table 
4.4.6.1.3-1, ``Reactor Vessel Material Surveillance Program--Withdrawal 
Schedule,'' with a note clarifying that surveillance capsule 
withdrawals are to be scheduled for the nearest vessel refueling outage 
date subsequent to the withdrawal time specified in the TS Table.
    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. The NRC staff has reviewed the licensees analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
Extended Use of Pressure-Temperature Limits
    The proposed change to the TSs to extend the use of the P-T limits 
does not affect the operation or configuration of any plant equipment. 
Thus, no new accident initiators are created by this change. The 
proposed change extends the use of the pressure-temperature (P-T) 
limits for an additional cycle of operation. The P-T curves prohibit 
operational conditions in which brittle fracture of the reactor vessel 
materials is possible. The P-T limits are based on the projected 
reactor vessel neutron fluence at 32 effective full power years (EFPY) 
of operation. At the end of the next cycle of operation, Cycle 10, 
Limerick Generating Station (LGS) Unit 1 will have attained a maximum 
of 48.1 percent of the 32 EFPY operating time which provides 
significant margin to ensure that the current 32 EFPY fluence 
projection will not be exceeded. This ensures that the basis for 
proposed applicability of the P-T limits is conservative and that the 
reactor vessel integrity is protected under all operating conditions. 
Therefore, neither the probability nor the consequences of an accident 
are increased.
Deferral of Withdrawal of Vessel Surveillance Specimens
    Deferring the withdrawal of the vessel surveillance capsules will 
not initiate or is not a precursor to any of the accident scenarios 
presented in the Updated Final Safety Analysis Report (UFSAR). This 
schedular adjustment will not increase the likelihood of equipment 
failure, will not defeat the design reactor protection functions, and 
will not increase the likelihood of failure of any plant structure, 
system or component. Therefore, neither the probability nor the 
consequences of an accident are increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
Extended Use of Pressure-Temperature Limits
    The proposed change to the technical specifications to extend the 
use of the P-T limits does not affect the operation or configuration of 
any plant equipment. The current P-T limits will remain valid and 
conservative during the proposed extension. Thus, no new or different 
accidents are created by this proposed change.
Deferral of Withdrawal of Vessel Surveillance Specimens
    The proposed deferral of the withdrawal of the vessel surveillance 
capsule does not involve a change to the plant design or operation. No 
new equipment will be installed or utilized, and no new operating 
conditions will be initiated as a result of this change. Because the P-
T limit curves are not impacted, the safety function of the reactor 
vessel to mitigate the release of radioactive steam and limit reactor 
inventory loss under normal, accident, and transient conditions is not 
affected. Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
Extended Use of Pressure-Temperature Limits
    The proposed change extends the use of the current P-T limits for 
an additional cycle of operation. No changes to the P-T limits are 
proposed. The current P-T limit curves are based on the projected 
reactor vessel neutron fluence after 32 EFPY of operation. At the end 
of the next operating cycle, Cycle 10, LGS Unit 1 will have attained a 
maximum of 48.1 percent of the 32 EFPY reactor vessel neutron fluence 
projection upon which the current P-T curves are based. The maximum 
operating time at the end of Cycle 10, when compared with the maximum 
operating time assumed for the P-T limits curves, ensures that the P-T 
limits will remain conservative and will ensure that the current 
margins for reactor pressure vessel integrity are unchanged. The 
proposed change maintains the relative margin of safety commensurate 
with that which existed at the time the American Society of Mechanical 
Engineers Boiler & Pressure Vessel Code, Section XI, Appendix G, was 
approved in 1974. No plant safety limits, setpoints, or design 
parameters are adversely affected by the proposed TS change. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.
Deferral of Withdrawal of Vessel Surveillance Specimens
    No plant safety limits, set points, or design parameters are 
adversely affected by the proposed deferral of withdrawal of vessel 
surveillance specimens. The deferral of the withdrawal of the vessel 
surveillance specimens does not affect the current P-T limit curves, 
and therefore, does not affect a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the

[[Page 59507]]

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: James W. Clifford.

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

    Date of amendment request: October 18, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for St. Lucie, Units 1 and 2, 
regarding Engineered Safety Feature Actuation System (ESFAS) 
instrumentation. Specifically, it would limit the period of time that 
inoperable recirculation actuation signal (RAS), containment spray 
actuation signal (CSAS), and auxiliary feedwater actuation signal 
(AFAS) input channels could be in the bypassed and/or tripped 
condition. Generally, the proposed TS employ a 48-hour completion time 
to restore an inoperable channel, which, in most cases, is more 
restrictive than the existing TS, and is comparable to the value used 
in the Standard TS for Combustion Engineering plants.
    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 operation of the facility in accordance with the 
proposed amendments involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No, facility operation under the new Technical Specification 
(TS) restrictions would not increase the probability of occurrence 
of any accident previously evaluated. The proposed changes only 
affect the ESFAS functions of RAS, CSAS, and AFAS; generally 
limiting the time that any instrument channel may be inoperable in a 
bypassed or tripped condition. No physical plant changes are 
proposed in conjunction with these revisions. The proposed changes 
to RAS and AFAS channel operability greatly reduce the time that 
actuation systems are vulnerable to spurious, inadvertent actuation. 
The proposed changes do allow a new unlimited time for trip of one 
CSAS channel on Unit 1. Although this increases the possibility of a 
spurious channel trip with a potential for causing an inadvertent 
spray actuation, this is offset by the increased reliability of 
spray in this configuration. Unit 2 already contains provision for 
the indefinite single channel trip of CSAS, and this change will 
also make the two units similar. Additionally, it is important to 
note that inadvertent actuation of any of these functions (RAS, 
CSAS, or AFAS) during plant operation is not an accident initiating 
event. Therefore, with no physical effects on the plant and no 
increase in probability that the subject ESFAS functions will 
initiate an accident, there is no increased probability that any 
previously evaluated accident will occur. The changes provided in 
this safety evaluation do not affect the assumptions or results of 
any accident evaluated in the UFSAR [Updated Final Safety Analysis 
Report].
    Likewise, the consequences of any accident previously evaluated 
have not been increased. The proposed changes, by limiting the time 
that ESFAS functions are inoperable, will increase the reliability 
of the associated ESFAS functions to respond to accidents. In 
particular, the revision to the RAS TS will limit the time that the 
RAS will be vulnerable to single failure and will therefore improve 
the system reliability during an accident. As these proposed changes 
constitute no physical change to the facility and only serve to 
increase ESF function reliability, FPL concludes that the 
consequences of previously evaluated accidents are not increased. 
The ability of the ESFAS to respond to accident conditions as 
assumed in any accident analysis has not been affected.
    (2) Would operation of the facility in accordance with the 
proposed amendments create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No, the proposed activity does not create the possibility of an 
accident of a different type than any previously evaluated. The 
proposed changes only affect the ESFAS functions of RAS, CSAS, and 
AFAS; generally limiting the time that any instrument channel may be 
inoperable in a bypassed or tripped condition. No physical plant 
changes are proposed in conjunction with these revisions. Thereby, 
the proposed changes do not create any new equipment interfaces, 
equipment response characteristics, or operating configurations. 
Without creation of a new interaction of materials, operating 
configuration, or operating interface, there is no possibility that 
the proposed changes can introduce a new or different kind of 
accident.
    (3) Would operation of the facility in accordance with the 
proposed amendments involve a significant reduction in a margin of 
safety?
    The margin of safety as defined in the basis for any Technical 
Specification or in any licensing document has not been reduced. The 
TS Bases for the associated ESF LCO [Limiting Condition for 
Operation] do not explicitly discuss a related margin of safety. 
However, by virtue of the increased ESFAS reliability provided by 
the proposed amendments, it is evident that the margin of safety 
will not be reduced in any manner.

    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: Richard P. Correia.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: October 17, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 6.8.4.h, ``Containment Leakage Rate 
Testing Program,'' to allow only one-time deviation from the 10-year 
frequency of the performance-based leakage rate testing program for 
Type A tests as recommended by Nuclear Energy Institute, NEI 94-01, 
Revision 0, ``Industry Guideline for Implementing Performance-Based 
Option of 10 CFR part 50, Appendix J,'' and endorsed by Regulatory 
Guide 1.163, ``Performance-Based Containment Leak-Rate Program.'' The 
one-time deviation would allow intergrated leak rate testing (ILRT) at 
no more than 15 years after the last ILRTs, performed in November 1992 
and October 1991 for Units 3 and 4 respectively.
    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 amendment to the Technical Specifications adds a 
one-time extension to the current interval for Type A (ILRT) 
testing. The current test interval of ten years, based on past 
performance, 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 testing extension is not a 
modification, nor a change to the operation of the plant, and the 
test extension is not a type that could lead to equipment failure or 
accident initiation. The proposed extension of Type A testing does 
not involve a significant increase in the consequences of an 
accident since research documented in NUREG-1493 has found that, 
generically, very few potential containment leakage paths are not 
identified with Type B and C tests. In fact, an analysis of 144 ILRT 
results, including 23 failures, found that no failures

[[Page 59508]]

were due to containment liner breech. The NUREG concluded that 
reducing the Type A frequency to one per twenty years was found to 
lead to an imperceptible increase in risk.
    Florida Power & Light provides a high degree of assurance 
through testing and inspection that the containment will not degrade 
in a manner detectable only by Type A testing. The last four Type A 
tests for both Turkey Point Units 3 and 4 show leakage rates well 
below acceptance criteria, indicating a leak-tight containment. 
Inspections required by the Maintenance Rule [10 CFR 50.65] and ASME 
[American Society of Mechanical Engineers] code, will identify 
indications of containment structure degradation that could affect 
that leak tightness. Type B and C testing required by Technical 
Specifications will identify any containment openings, such as 
valves, that would otherwise be detected by the Type A tests. These 
factors show that the Turkey Point Units 3 and 4 Type A test 
extension will not represent a significant increase in the 
consequences of an accident.
    Based on the above, it is concluded that the proposed amendments 
to extend the Type A test frequency does not involve a significant 
increase in the probability or consequences of any 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 previously evaluated.
    The proposed change does not create a new or different type of 
accident for Turkey Point because no physical plant changes are 
being made, and no compensatory measures are imposed that would 
create a new failure scenario. The proposed change only requests a 
one-time extension to the current interval for Type A testing. The 
current test interval of 10 years, based on past performance, 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 create the 
possibility of a new or different type of accident because there are 
no physical changes being made to the plant, and there are no 
changes to the operation of the plant that could introduce a new 
failure mode creating an accident or affecting the mitigation of an 
accident.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed license amendment requests a one-time extension to 
the current interval for Type A testing. The current test interval 
of ten years, based on past performance, would be extended on a one-
time basis to 15 years from the last Type A test. The proposed 
extension to Type A testing will not significantly reduce the margin 
of safety. The NUREG-1493 generic study of the effects of extending 
containment leakage testing found that a 20-year test interval for 
Type A leakage testing resulted in an imperceptible increase in risk 
to the public. NUREG-1493 found that, generically, the design 
containment leakage rate contributed about 0.1 percent to the 
individual risk and that the decrease in Type A testing frequency 
would have minimal effect on this risk, since 95 percent of the 
potential leakage paths are detected by Type B and C testing. A 
Turkey Point plant-specific risk calculation is consistent with the 
generic conclusions identified in NUREG-1493.
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the plant's margin of 
safety.

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

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

    Date of amendment requests: August 7, 2001.
    Description of amendment requests: The proposed amendments would 
create Technical Specification (TS) 3.0.6 and associated bases to allow 
equipment that was removed from service or declared inoperable to be 
returned to service under administrative controls solely to perform the 
testing required to demonstrate its operability or the operability of 
other equipment. TS 3.0.6 would incorporate the administrative controls 
currently approved for use as TS 3.0.5 in NUREG-1431, ``Standard 
Technical Specifications Westinghouse Plants,'' Revision 2, dated April 
30, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Does the change involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated?

Probability of Occurrence of an Accident Previously Evaluated

    The potential impact of temporarily returning the equipment to 
service is considered to be insignificant since the equipment will 
either be expected to be able to perform its required safety 
function or sufficient redundancy will exist such that the function 
would still occur if required. This is addressed in Generic Letter 
(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical 
Specifications (STS) on the Applicability of Limiting Conditions for 
Operation and Surveillance Requirements.'' GL 87-09 states, ``It is 
overly conservative to assume that systems or components are 
inoperable when a surveillance has not been performed because the 
vast majority of surveillances do in fact demonstrate that systems 
or components are operable.'' In addition, returning the equipment 
to service for testing will promote timely restoration of the 
equipment. Therefore, the proposed changes do not significantly 
affect accident initiators or precursors.
    The proposed change to create a Bases statement for TS 3.0.6 
provides explanatory information regarding the intent of the 
specification and how it is to be implemented. The proposed Bases 
change does not alter requirements of the associated TS. Therefore, 
the effect of the Bases change on accident initiators and precursors 
of an accident is bounded by the effect of the TS change as 
described above. The format changes are intended to improve 
appearance and do not alter any requirements.
    Therefore, the proposed changes do not adversely affect any 
accident initiators or precursors and will not involve a significant 
increase in the probability of an accident previously evaluated.

Consequences of an Accident Previously Evaluated

    The proposed change will allow temporarily returning equipment, 
that was previously declared inoperable, to service in a state in 
which it is expected to function to mitigate the consequences of a 
previously analyzed accident. The proposed change will also permit 
temporarily restoring inoperable equipment to service in situations 
where sufficient redundancy would exist for its function to mitigate 
the consequences of a previously analyzed accident to be performed. 
This will promote timely restoration of equipment and capabilities 
to mitigate the consequences of an accident previously analyzed.
    The proposed change to include a Bases statement for TS 3.0.6 
provides explanatory information regarding the intent of the 
specification and how it is to be implemented. The proposed Bases 
change does not alter requirements of the associated TS. Therefore, 
the effect of the Bases change on offsite dose consequences of an 
accident previously analyzed is bounded by the effect of the TS 
change as described above. The format changes are intended to 
improve appearance and do not alter any requirements.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not significantly increased.

2. Does the change create the possibility of a new or different kind of 
accident from any accident previously evaluated?

    The proposed changes do not introduce a new mode of plant 
operation and do not involve a physical modification to the plant. 
Operation with the inoperable equipment temporarily restored to 
service under administrative controls is not considered a new mode 
of operation since the equipment is not being physically altered. As 
such, the manner in which it can fail remains the same.

[[Page 59509]]

    The proposed change to include a Bases statement for TS 3.0.6 
provides explanatory information regarding the intent of the 
specification and how it is to be implemented. The proposed Bases 
change does not alter requirements of the associated TS. Therefore, 
the effect of the Bases changes on accident initiators or precursors 
is bounded by the effect of the associated TS as described above. 
The format changes are intended to improve appearance and do not 
alter any requirements.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

3. Does the change involve a significant reduction in a margin of 
safety?

    The proposed new TS 3.0.6 can be applied to any structures, 
systems, and components that are governed by the TS. As such, the 
proposed changes are applicable to every margin of safety imposed by 
the TS.
    The proposed change will allow temporarily returning equipment 
that was previously declared inoperable to service in a state in 
which it is expected to function to mitigate the consequences of a 
previously analyzed accident. The proposed change will also permit 
temporarily restoring inoperable equipment to service in situations 
where sufficient redundancy would exist for its function to mitigate 
the consequences of a previously analyzed accident to be performed. 
The performance of the testing should confirm the expected 
capability of the equipment and there is no significant impact on 
any TS safety setting or setpoint.
    There is no margin of safety pertinent to the proposed Bases 
change. The format changes are intended to improve appearance and do 
not alter any requirements.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety. In summary, based upon the above 
evaluation, I&M has concluded that the proposed amendment involves 
no significant hazards consideration.

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

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

    Date of amendment request: May 9, 2001.
    Description of amendment request: The proposed amendment would 
change the Technical Specification (TS) to correct an error in TS Table 
3.3.1.1-1 Function 2.b, correct a typographical error in labeling 
surveillance requirement 3.3.1.1.13, and revise bases pages B 3.3-8 and 
B 3.3-10.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change is to correct an error in documentation that was 
introduced during implementation of Amendment 151 and retained in TS 
during the conversion to ITS [Improved Technical Specifications] as 
well as an error that was introduced into TS during the conversion 
to ITS. Neither the design basis nor the functionality of the 
instrumentation is being physically changed. The Neutron Monitoring 
system performs a mitigating function and is not an accident 
initiating system. The actual mitigating function of the Neutron 
Monitoring is not changed. Only an implied but non-existent 
mitigating capability is being removed from TS. This change does not 
create or modify any accident initiators. Therefore, there is no 
increase in the probability of an accident previously evaluated.
    The APRM [Average Power Range Monitor] system is credited for 
mitigating the consequences of the Control Rod Drop Accident. The 
APRM system also provides protection for the reactor to mitigate the 
consequences of such abnormal operational transients as loss of 
feedwater heater, pressure regulator failure, or Main Steam 
Isolation Valve closure. The proposed change will not change the 
functionality or setpoints for either the APRM Flux-High (Fixed) or 
the APRM Flux-High (Biased) functions. Additionally, the correction 
of an incorrect Surveillance Requirement reference does not change 
how any surveillance is performed. Therefore the consequences of an 
accident previously evaluated will not be increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Since this change in the TS does not involve a physical change 
to the instrumentation, to the setpoints, or to the design or 
functionality of the circuitry for reactor scram on APRM Flux-High, 
fixed or flow-biased, the change does not create a possibility of a 
new or different kind of accident not previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The setpoints for the Neutron Flux-High instrumentation are not 
changed by this proposed TS change. The safety function allowable 
value setpoint remains at less than or equal to 120% RTP [rated 
thermal power]. The formula for the APRM Flux-High (flow biased) is 
not being changed. Since neither of these is being changed, the 
margin of safety is not reduced.

    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. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: August 6, 2001, as supplemented November 
2, 2001.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station Technical Specifications (TS) Index, TS 
1.0, ``Definitions,'' and TS Table 1.2, ``Operational Modes,'' to 
reflect the improved Standard Technical Specifications for Westinghouse 
plants.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to TS Index, TS 1.0 and TS Table 1.2 are 
changes that do not change any structures, systems or components 
(SSCs) thus, the proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility. In addition, the 
proposed changes do not affect the manner in which the plant 
responds in normal operation, transient or accident conditions. The 
proposed changes do not alter or prevent the ability of SSCs to 
perform their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR). Finally, while these changes 
may afford North Atlantic operational flexibility, the changes are 
an enhancement and do not affect plant safety.
    The proposed changes do not affect the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated in the 
Seabrook Station UFSAR. Further, the proposed changes do not 
increase the types and amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures.
    Therefore, it is concluded that these proposed revisions to TS 
Index, TS 1.0 and TS Table 1.2 do not involve a significant increase 
in the probability or consequence of an accident previously 
evaluated.

[[Page 59510]]

    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    This [sic] proposed changes to TS Index, TS 1.0 and TS Table 1.2 
are changes that do not change the operation or the design basis of 
any plant system or component during normal or accident conditions. 
The proposed change incorporates definitions delineated in the 
improved Standard Technical Specifications (NUREG-1431). The 
proposed changes do not include any physical changes to the plant. 
In addition, the proposed changes do not change the function or 
operation of plant equipment or introduce any new failure 
mechanisms. The plant equipment will continue to respond per the 
design and analyses and there will not be a malfunction of a new or 
different type introduced by the proposed changes.
    The proposed changes are administrative in nature and only 
correct, update and clarify the Seabrook Station Operating License 
to reflect the definitions in the improved Standard Technical 
Specifications. The proposed changes do not modify the facility nor 
do they affect the plant's response to normal, transient or accident 
conditions. The changes do not introduce a new mode of plant 
operation. While these changes may afford North Atlantic operational 
flexibility, the changes are an enhancement and do not affect plant 
safety. The plant's design and design basis are not revised and the 
current safety analyses remains in effect.
    Thus, these proposed revisions to TS Index, TS 1.0 and TS Table 
1.2 do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes to TS Index, TS 1.0 and TS Table 1.2 are 
administrative in nature and only correct, update and clarify the 
Seabrook Station Operating License to reflect the improved Standard 
Technical Specifications. While these changes may afford North 
Atlantic operational flexibility, the changes are an enhancement and 
do not affect plant safety. The safety margins established through 
Limiting Conditions for Operation, Limiting Safety System Settings 
and Safety Limits as specified in the Technical Specifications are 
not revised nor is the plant design revised by the proposed changes.
    Thus, it is concluded that these proposed revisions to TS Index, 
TS 1.0 and TS Table 1.2 do not involve a significant reduction in a 
margin of safety.
    Based on the above evaluation, North Atlantic concludes that the 
proposed changes to TS Index, TS 1.0 and TS Table 1.2 do not 
constitute a significant hazard.

    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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: November 2, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Table 3.3.1-1, Item 1, ``Variable 
High Power Trip'' (VHPT), by increasing the maximum allowable value for 
the VHPT from 106.5 percent to 111 percent.
    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:

    Nuclear Management Company has evaluated whether or not a 
significant hazards consideration is involved with the proposed 
amendment by focusing on the three standards set forth in 10 CFR 
50.92, ``Issuance of Amendment.'' The following evaluation supports 
the finding that operation of the facility in accordance with the 
proposed change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the maximum Allowable Value for the 
Variable High Power Trip (VHPT) function in the Technical 
Specifications would not change or remove any considerations of 
uncertainties from the FSAR [Final Safety Analysis Report] Chapter 
14 Safety Analysis. The methodology that was utilized in determining 
the recommended change in the maximum allowable value follows 
standard ANSI/ISA-S67.04-1994, ``Setpoints for Nuclear Safety-
Related Instrumentation,'' and NRC Regulatory Guide 1.105, 
``Setpoints for Safety-Related Instrumentation,'' Revision 3. With 
the proposed changes to the maximum allowable value and calculated 
setpoint of the VHPT in place, the reactor is still protected from 
reaching the analytical limit of 115% reactor power.
    Therefore, operation of the facility in accordance with the 
proposed change to the Technical Specifications would 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 previously evaluated.
    The proposed changes to the maximum Allowable Value and 
Calculated Setpoint for the Variable High Power Trip function in the 
Technical Specifications would not change or add a system function. 
The proposed change alters the way the uncertainties (including 
uncertainties of instrument measurement and calibration) are 
accounted for without actually removing uncertainties from the 
calculation. This proposed change follows the standard ANSI/ISA-
S67.04-1994 and NRC Regulatory Guide 1.105, Revision 3.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the maximum Allowable Value for the 
Variable High Power Trip function in the Technical Specifications 
would account for all uncertainties in the VHP trip setpoint 
calculation, instead of taking them into account in the maximum 
allowable value calculation, as is currently done. In addition, 
double accounting for nuclear instrumentation uncertainties has been 
removed. The uncertainties will still be taken into account in 
determining the calculated setpoint based on the maximum allowable 
value of the VHPT, in accordance with the standard ANSI/ISA-S67.04-
1994 and NRC Regulatory Guide 1.105, Revision 3. This methodology 
continues to assure that the Analytical Limit will not be exceeded.
    Therefore, the proposed change to the Technical Specifications 
would not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based upon 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: William D. Reckley (Acting).

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 amendment request: November 1, 2001.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TSs) to allow a one-time extension 
of the allowed outage time for the control room emergency filtration 
system (CREFS) from 7 days to 30 days. The licensee is requesting this 
one-time change in order to implement modifications to the CREFS.
    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 Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.

[[Page 59511]]

    The operability of CREFS ensures that the control room will 
remain habitable for operators during and following all credible 
accident conditions. The inoperability or failure of CREFS is not an 
accident initiator or precursor. Therefore, the probability of an 
accident previously evaluated will not be significantly increased as 
a result of the proposed change. Because design limitations continue 
to be met and the integrity of the reactor coolant system pressure 
boundary is not challenged, the assumptions employed in the 
calculation of the offsite radiological doses remain valid. Control 
room dose calculations are not affected outside the limited one-time 
period when the CREFS modifications/upgrades are ongoing.
    During the period that CREFS will be inoperable, temporary 
ventilation will provide adequate filtration to the control room and 
adequate cooling to the control and computer rooms. The 
effectiveness of the temporary filtration provided during this 30 
day period is not significantly less than that of the permanently 
installed CREFS. Only the duration of a currently allowed outage 
time is being changed, with commensurate compensatory measures being 
taken. Therefore, the consequences of an accident previously 
evaluated will not be significantly increased as a result of the 
proposed change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created as a result of this 
amendment. The evaluation of the effects of the proposed changes 
indicate that all design standards and applicable safety criteria 
limits are met. These changes therefore do not cause the initiation 
of any new or different accident nor create any new failure 
mechanisms.
    Equipment important to safety will continue to operate as 
designed. Only the duration of a system's allowed outage time is 
being changed. Component integrity is not challenged. The changes do 
not result in any event previously deemed incredible being made 
credible. The changes do not result in more adverse conditions or 
result in any increase in the challenges to safety systems. 
Therefore, operation of the Point Beach Nuclear Plant in accordance 
with the proposed amendments will not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The CREFS functions to mitigate the effects of accidents. 
Implementation of the modifications/upgrades will require removing 
the system from service for a period of time longer than presently 
allowed by the Technical Specification. This results in a longer 
period during which the consequences of a design basis accident, 
affecting the dose of control room personnel, may be slightly 
increased. During the period that CREFS will be inoperable, a 
temporary ventilation system will provide adequate filtration to the 
control room and adequate cooling to the control and computer rooms. 
The effectiveness of the temporary filtration provided during this 
30 day period is not significantly less than that of the permanently 
installed CREFS. Only the duration of a currently allowed outage 
time is being changed, with commensurate compensatory measures being 
taken. There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. The 
proposed modification will not otherwise affect the plant protective 
boundaries, will not cause a release of fission products to the 
public, nor will it degrade the performance of any other SSCs 
important to safety. The analysis for the limiting design basis 
accident, the large break LOCA, has a significant amount of 
conservatism built in to account for uncertainties in system 
performance an analysis techniques. This conservative margin of 
safety, along with the temporary filtration unit, provide a high 
level of confidence that the health and safety of the operators will 
be maintained, such that they will be able to prevent or mitigate an 
event. Therefore, removing the CREFS from service for 30 days on a 
one-time basis to permit system upgrading, will not significantly 
reduce the margin of safety. The improvements to CREFS resulting 
from the proposed modifications will enhance operator protection 
against conditions resulting from a design basis accident and 
therefore provide a net benefit to radiological health and reactor 
safety.

Conclusion

    Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of any accident previously analyzed; 
will not result in a new or different kind of accident from any 
accident previously analyzed; and, does not result in a significant 
reduction in any margin of safety. Therefore, operation of PBNP 
[Point Beach Nuclear Plant] in accordance with the proposed 
amendments does not result in a significant hazards determination.

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

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

    Date of amendment request: October 18, 2001.
    Description of amendment request: The proposed amendments would 
modify the Technical Specification Surveillance Requirement (SR) 
3.4.3.1 for testing of the main steam safety relief valves (MSRVs) so 
that the setpoint tolerance for ``As-Found'' testing would be changed 
from 1 percent to  3 percent. The requirements 
for testing of the tolerances associated with ``As-left'' testing would 
remain unchanged. An editorial change would also be made to remove a 
note regarding an associated relief request.
    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 action does not involve a significant increase 
in the probability or consequences of an accident as previously 
evaluated.
    The proposed change allows an increase in the as-found MSRV 
safety mode setpoint tolerance, determined by test after the valves 
have been removed from service, from 1% to 
3%. The proposed change does not alter the TS 3.4.3 
Surveillance Requirements on the nominal MSRV safety mode lift 
setpoints, the MSRV relief mode setpoints, the required frequency 
for the MSRV lift setpoint tests, or the number of MSRVs required to 
be operable.
    Consistent with current requirements, this change continues to 
require that these valves be adjusted to within 1% of 
their nominal lift setpoints following testing. The proposed action 
does not change any other behavior or operation of any MSRV, and 
therefore, has no significant impact on the reactor operation. It 
also has no significant impact on response to any perturbation of 
reactor operation including transients and accidents previously 
analyzed in the Final Safety Analysis Report (FSAR).
    The proposed action does not involve physical changes to the 
valves, nor does it change the safety function of the valves. The 
proposed TS revision involves no significant changes to the 
operation of any systems or components in normal or accident 
operating conditions and no changes to existing structures, systems, 
or components. Therefore, these changes will not increase the 
probability of an accident previously evaluated.
    Generic considerations related to the change in setpoint 
tolerance were addressed in NEDC-31753P, ``BWROG In-Service Pressure 
Relief Technical Specification Revision Licensing Topical Report,'' 
and were reviewed and approved by the NRC in a Safety Evaluation 
Report (SER) dated March 8, 1993. The plant specific evaluations, 
required by the NRC's SER and performed to support this proposed 
change, show that there is adequate margin to the design core 
thermal limits and to the reactor vessel pressure limits using a 
3% setpoint tolerance. These analyses also show that

[[Page 59512]]

operation of the high pressure coolant injection (HPCI) and reactor 
core isolation cooling (RCIC) systems are not adversely affected and 
the containment response from a loss of coolant accident is 
acceptable. The plant systems associated with these proposed changes 
are capable of meeting all applicable design basis requirements and 
retain the capability to mitigate the consequences of accidents 
described in the FSAR. Therefore, these changes do not involve an 
increase in the consequences of any accident previously evaluated.
    Therefore, the proposed amendment does not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed action does not create a possibility of a new or 
different kind of accident than previously evaluated.
    The proposed change was developed in accordance with the 
provisions contained in the NRC SER, dated March 8, 1993, for the 
``BWR Owners Group Inservice Pressure Relief Technical Specification 
Revision Licensing Topical Report,'' NEDC-31753P. The revised MSRV 
setpoint tolerance limit does not adversely impact the operation of 
any safety-related component or equipment. Since the proposed action 
does not involve hardware changes, significant changes to the 
operation of any systems or components, nor changes to existing 
structures, systems, or components, there is no possibility that a 
new or different kind of accident is created.
    The proposed change to allow an increase in the MSRV safety mode 
setpoint tolerance from 1% to 3% does not 
alter the nominal MSRV lift setpoints or the number of MSRVs 
currently required to be operable by SSES Technical Specifications. 
The proposed action does not involve physical changes to the valves, 
nor does it change the safety function of the valves. The proposed 
action does not involve a physical alteration of any existing plant 
equipment. No new or different equipment is being installed. There 
is no alteration to the parameters within which the plant is 
normally operated. As a result no new failure modes are being 
introduced. There are no changes in the procedures governing normal 
plant operation, nor the procedures utilized to respond to plant 
transients.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed action does not involve a significant reduction 
in a margin of safety.
    The proposed action does not involve a significant reduction in 
a margin of safety. Establishment of the 3% MSRV safety 
setpoint tolerance limit does not adversely impact the operation of 
any safety-related component or equipment. Engineering evaluations 
concluded that there are no significant impacts on fuel thermal 
limits, safety related systems, structures or components, and no 
significant impact on the accident analyses associated with the 
proposed changes.
    The margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change does not significantly impact the condition or 
performance of structures, systems, and components relied upon for 
accident mitigation.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Lakshminaras Raghaven.

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

    Date of amendment request: September 20, 2001.
    Description of amendment request: The proposed amendments would 
support extension of the operating cycle from 18 months to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    a. Surveillance Testing Interval Extensions.
    The proposed Technical Specification (TS) change involves a 
change in the surveillance testing intervals to facilitate a change 
in the operating cycle from 18 months to 24 months. The proposed TS 
change does not physically impact the plant, nor does it impact any 
design or functional requirements of the associated systems. That 
is, the proposed TS change neither degrades the performance of, nor 
increases the challenges to, any safety systems assumed to function 
in the plant safety analysis. The proposed TS change neither impacts 
the TS SRs [surveillance requirements] themselves nor the manner in 
which the surveillances are performed.
    In addition, the proposed TS change does not introduce any 
accident initiators, since no accidents previously evaluated relate 
to the frequency of surveillance testing. Also, evaluation of the 
proposed TS change demonstrates that the availability of equipment 
and systems required to prevent or mitigate the radiological 
consequences of an accident is not significantly affected because of 
other, more frequent testing that is performed, the availability of 
redundant systems and equipment, or the high reliability of the 
equipment. Since the impact on the systems is minimal, it is 
concluded the overall impact on the safety analysis is negligible.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicate there is no evidence of 
any failure that would invalidate the above conclusions. Therefore, 
the proposed TS change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    b. Allowable Value Changes.
    A change in Allowable Values is proposed for Table 3.3.5.1-1, 
Item 2.f. The proposed change is the result of application for the 
Hatch Instrument Setpoint Methodology using plant-specific drift 
values. Application of this methodology results in Allowable Values 
that more accurately reflect total instrumentation loop accuracy, as 
well as that of test equipment and calculated drift between 
surveillances. The proposed change will not result in any hardware 
changes. The instrumentation is not assumed to be an initiator of 
any analyzed event. Existing operating margin between plant 
conditions and actual plant setpoints is not significantly reduced 
due to the proposed changes. The role of the instrumentation is in 
mitigating and thereby, limiting the consequences of accidents.
    The Allowable Values were developed to ensure the design and 
safety analysis limits are satisfied. The methodology used for the 
development of the Allowable Values ensures: 1) the affected 
instrumentation remains capable of mitigating design basis events as 
described in the safety analysis and 2) the results and radiological 
consequences described in the safety analysis remain bounding. 
Additionally, the proposed change does not alter the plant's ability 
to detect and mitigate events. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    c. Surveillance Testing Interval Reduction to Semiannual.
    The proposed TS change involves a reduction in the surveillance 
testing interval from 18 months to 184 days for the instrumentation 
associated with Table 3.3.8.2-1. The shorter intervals are based 
upon the plant-specific results of a review of the surveillance test 
history for the devices. The implementing procedures for these SRs 
have been performed on a 184-day interval for a number of years, and 
this change more accurately reflects actual plant maintenance 
practices. The proposed, more restrictive TS change does not 
physically impact the plant, nor does it impact any design or 
functional requirements of the associated systems. That is, the 
proposed TS change neither degrades the performance of, nor 
increases the challenges to, any safety systems assumed to function 
in the safety analysis. This proposed TS change neither impacts the 
TS SRs

[[Page 59513]]

themselves nor the manner in which the surveillances are performed.
    In addition, the proposed TS change does not introduce any 
accident initiators, since no accidents previously evaluated relate 
to the frequency of surveillance testing. The proposed TS intervals 
demonstrate that the equipment and systems required to prevent or 
mitigate the radiological consequences of an accident are continuing 
to meet the assumptions of the setpoint evaluation on a more 
frequent basis. Since the impact on the systems is minimal, and the 
assumptions of the safety analyses are maintained, it is concluded 
the overall impact on the plant safety analysis is negligible.
    Furthermore, setpoint drift evaluations prepared for the subject 
instrumentation show that the existing Allowable Values are 
acceptable without change. Therefore, the proposed TS change does 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    d. Change of CHANNEL CALIBRATION to CHANNEL FUNCTIONAL TEST for 
Float Switches.
    The proposed TS change involves a change in the SRs from CHANNEL 
CALIBRATIONS to CHANNEL FUNCTIONAL TESTS for float switches used in 
Table 3.3.1.1-1, Item 7.b; Table 3.3.5.1-1, Item 3.d; and Table 
3.3.5.2-1, Items 3 and 4. The float switches are mechanical devices 
that require mechanical setting at the proper level only. Because 
the devices cannot be significantly adjusted without a physical 
change in the location of the installation, the CHANNEL FUNCTIONAL 
TEST provides all the functionality of a CHANNEL CALIBRATION for 
this type of device. Therefore, the change in type of SR does not 
impact the actual testing requirements for the subject devices.
    The proposed TS change does not physically impact the plant, nor 
does it impact any design or functional requirements of the 
associated systems. That is, the proposed TS change neither degrades 
the performance of, nor increases the challenges to, any safety 
systems assumed to function in the safety analysis. The proposed TS 
change does not impact the manner in which the surveillances are 
performed.
    In addition, the proposed TS change does not introduce any 
accident initiators, since the same functional requirements exist 
with the proposed change. Also, evaluation of the proposed TS change 
demonstrates the availability of equipment and systems required to 
prevent or mitigate the radiological consequences of an accident is 
not significantly affected because of the availability of redundant 
systems and equipment and the high reliability of the equipment. 
Since the impact on the systems is minimal, it is concluded the 
overall impact on the plant safety analysis is negligible.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicated that there was no 
evidence of any failures that would invalidate the above 
conclusions. Therefore, the proposed TS change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    a. Surveillance Testing Interval Extensions.
    The proposed TS change involves a change in the surveillance 
testing intervals to facilitate a change in the operating cycle 
length. The proposed TS change does not introduce any failure 
mechanisms of a different type than those previously evaluated, 
since there are no physical changes being made to the facility. No 
new or different equipment is being installed. No installed 
equipment is being operated in a different manner. As a result, no 
new failure modes are introduced. In addition, the SRs themselves, 
and the manner in which surveillance tests are performed, remain 
unchanged.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicate there is no evidence of 
any failure that would invalidate the above conclusions. Therefore, 
the proposed TS change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    b. Allowable Value Changes.
    The proposed change in Allowable Values is the result of 
application of the Instrument Setpoint Methodology using plant-
specific drift values and does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. This is based upon the fact that the method and manner of 
plant operation are unchanged.
    The use of the proposed Allowable Values does not impact safe 
operation of the plant in that the safety analysis limits are 
maintained. The proposed change in Allowable Values involves no 
system additions or physical modifications to plant systems. The 
Allowable Values are revised to ensure the affected instrumentation 
remains capable of mitigating accidents and transients. Plant 
equipment will not be operated in a manner different from previous 
operation, except that setpoints may be changed. Since operational 
methods remain unchanged and the operating parameters were evaluated 
to maintain the plant within existing design basis criteria, no 
different type of failure or accident is created.
    c. Surveillance Testing Interval Reductions to Semiannual.
    The proposed TS change involves a change in the surveillance 
testing interval due to the review of the surveillance test history 
of the subject devices. Also, the semiannual tests reflect current 
HNP calibration practices. The proposed TS change does not introduce 
any failure mechanism of a different type than those previously 
evaluated, since the proposed change makes no physical changes to 
the plant. No new or different equipment is being installed. No 
installed equipment is being operated in a different manner.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicate there is no evidence of 
any failure that would invalidate the above conclusions. Therefore, 
the proposed TS change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    d. Change of CHANNEL CALIBRATION to CHANNEL FUNCTIONAL TEST for 
Float Switches.
    The proposed TS change does not impact the actual testing 
requirements for the subject devices. The proposed TS change does 
not introduce any failure mechanism of a different type than those 
previously evaluated, since the proposed change makes no physical 
changes to the plant. No new or different equipment is being 
installed. No installed equipment is being operated in a different 
manner. As a result, no new failure mode is being introduced. In 
addition, the SRs themselves, and the manner in which surveillance 
tests are performed, remain unchanged.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicates there is no evidence of 
any failure that would invalidate the above conclusions. Therefore, 
the proposed TS change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety
    a. Surveillance Testing Interval Extensions.
    Although the proposed TS change results in changes in the 
interval between surveillance tests, the impact, if any, on system 
availability is minimal, based upon other, more frequent testing 
that is performed, the existence of redundant systems and equipment, 
or overall system reliability. Evaluations show there is no evidence 
of any time-dependent failure that would impact the system 
availability.
    The proposed change does not significantly impact the condition 
or performance of structures, systems, and components relied upon 
for accident mitigation. The proposed change does not significantly 
impact any safety analysis assumptions or results. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    b. Allowable Value Changes.
    The proposed change does not involve a reduction in a margin of 
safety. The proposed change was developed using a methodology to 
ensure safety analysis limits are not exceeded. As such, this 
proposed change does not involve a significant reduction in a margin 
of safety.
    c. Surveillance Testing Interval Reductions to Semiannual.
    The proposed TS change results in a shorter interval between 
surveillance tests to ensure the assumptions of the safety analysis 
are maintained. The impact, if any, on system availability is 
minimal, as a result of the more frequent testing that is performed. 
The proposed change does not significantly impact the condition or 
performance of structures, systems, and components relied upon for 
accident mitigation. The proposed change does not significantly 
impact any safety analysis assumptions or results. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    d. Change of CHANNEL CALIBRATION to CHANNEL FUNCTIONAL TEST for 
Float Switches.

[[Page 59514]]

    The proposed TS change does not impact the actual testing 
requirements for the subject devices. The impact, if any, on system 
availability due to this change is minimal, based upon the existence 
of redundant systems and equipment and overall system reliability.
    An historical review of surveillance test results and associated 
maintenance records indicates there is no evidence of any failure 
that would invalidate the above conclusions. The proposed change 
does not significantly impact the condition or performance of 
structures, systems, and components relied upon for accident 
mitigation. The proposed change does not significantly impact any 
safety analysis assumptions or results. 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: September 20, 2001.
    Description of amendment request: The proposed amendments would 
change specified surveillances from 92 days to 184 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed Technical Specifications (TS) change involves an 
increase in the surveillance testing intervals for various 
Surveillance Requirements (SRs) from 92 days to 184 days. The 
proposed TS changes do not physically impact the plant, nor do they 
impact any design or functional requirements of the associated 
systems. That is, the proposed TS change does not degrade the 
performance of, or increase the challenges to, any safety systems 
assumed to function in the safety analysis. The proposed TS changes 
neither impact the TS SRs themselves nor the way in which the 
surveillances are performed. In addition, the proposed TS change 
does not introduce any accident initiators, since no accidents 
previously evaluated relate to the frequency of surveillance 
testing. Also, evaluation of the proposed TS change demonstrates 
that the availability of equipment and systems required to prevent 
or mitigate the radiological consequences of an accident are not 
significantly affected because of other, more frequent testing that 
is performed, the availability of redundant systems and equipment, 
or the high reliability of the equipment. Since the impact on the 
systems is minimal, it is concluded that the overall impact on the 
plant safety analysis is negligible.
    A sensitivity analysis was performed to determine the effect of 
the increased surveillance intervals on the HNP [Hatch Nuclear 
Plant] Probabilistic Risk Assessment (PRA). This sensitivity 
analysis shows a negligible increase in core damage frequency (CDF) 
and essentially no change in large early release frequency (LERF) 
due to the proposed change.
    Furthermore, an historical review of surveillance test results 
and associated maintenance record indicates there is no evidence of 
any failure that would invalidate the above conclusions. Therefore, 
the proposed TS change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change involves a change in the various SR 
intervals from 92 days to 184 days. The proposed TS change does not 
introduce any failure mechanisms of a different type than those 
previously evaluated, since no physical changes to the plant are 
being made. Also, no new or different equipment is being installed, 
and no installed equipment is being operated in a different manner. 
As a result, no new failure modes are introduced. In addition, the 
surveillance test requirements themselves, and the way surveillance 
tests are performed, remain unchanged.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicates there is no evidence of 
any failure that would invalidate the above conclusions. Therefore, 
the proposed TS change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    Although the proposed TS change results in changes to the 
interval between surveillance tests, the impact, if any, on system 
availability is minimal, based upon other, more frequent testing 
that is performed, the existence of redundant systems and equipment, 
or overall system reliability. Evaluations show there is no evidence 
of time-dependent failures that would impact the availability of the 
systems. The proposed change does not significantly impact the 
condition or performance of structures, systems, and components 
relied upon for accident mitigation.
    A sensitivity analysis was performed to determine the effect of 
the increased surveillance intervals on the HNP PRA. This 
sensitivity analysis shows a negligible increase in CDF and 
essentially no change in LERF due to the proposed change.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicates there was no evidence 
of any failure that would invalidate the above conclusions. 
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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Richard J. Laufer, Acting.

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

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

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: October 16, 2001.
    Brief description of amendment request: The proposed amendment 
would add a condition to the Operating License to extend certain 
Technical Specification surveillance requirement (SR) intervals, one 
time. The SR intervals would be extended up to 65 days, but no later 
than April 30, 2003, to permit them to be performed during

[[Page 59515]]

the next refueling outage, which has been rescheduled because the plant 
is currently in a forced extended outage.
    Date of publication of individual notice in Federal Register: 
November 13, 2001 (66 FR 56865).
    Expiration date of individual notice: December 13, 2001.

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/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

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

    Date of application for amendment: December 29, 2000, as 
supplemented March 22 and July 27, 2001.
    Brief description of amendment: The amendment increases the allowed 
outage time from 3 to 14 days for a single inoperable Division 1 or 2 
diesel generator.
    Date of issuance: November 8, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 141.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7668). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 2, 2001, as supplemented 
July 18, 2001.
    Brief description of amendment: The amendment extends the 
surveillance test interval of the slave relays of the Engineered Safety 
Features Actuation System from 90 days to 8 months.
    Date of issuance: November 5, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 198.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36337).
    The July 18, 2001, supplement was within the scope of the original 
application and did not change the staff's proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 2, 2001, as supplemented by 
letter dated August 23, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to not require the moderator temperature 
coefficient (MTC) determination in TS 4.1.1.4.2c if the results of the 
MTC determination required in TSs 4.1.1.4.2a and 4.1.1.4.2b are within 
a certain tolerance of the corresponding design values.
    Date of issuance: November 16, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 236.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31706).
    The August 23, 2001, supplemental letter provided clarifying 
information that was within the scope of the original Federal Register 
notice and did not change the staff's initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 16, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear 
Power Station, Unit 2, Grundy County, Illinois

    Date of application for amendment: June 6, 2001, as supplemented by 
letter dated September 17, 2001.
    Brief description of amendment: The amendment revises the values of 
the Safety Limit for the Minimum Critical Power Ratio in Technical 
Specification Section 2.1.1.
    Date of issuance: November 2, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 189.
    Facility Operating License No. DPR-19: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46479).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 2001.
    No significant hazards consideration comments received: No.

[[Page 59516]]

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

    Date of application for amendments: September 29, 2000, as 
supplemented by letters dated March 1, July 13, August 9, August 13, 
and October 17, 2001
    Brief description of amendments:The amendments change the technical 
specifications to reflect a change in fuel vendors from Siemens Power 
Corporation to General Electric, and a transition to GE14 fuel. As part 
of the transition, changes are made to the number of required automatic 
depressurization system valves and to the time delay relay settings on 
emergency core cooling system pumps. These changes were noticed in the 
Federal Register on December 27, 2000 (65 FR 81908), August 22, 2001 
(66 FR 44170), and August 23, 2001 (66 FR 44382).
    Date of issuance: November 2, 2001
    Effective date: As of the date of issuance and shall be implemented 
following refueling outage 17.
    Amendment Nos.: 188 and 183
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 30, 2001, as supplemented 
September 10, 2001.
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) Surveillance Requirement (SR) 3.6.1.1.3 
and adds two new SRs, SR 3.6.1.1.4 and SR 3.6.1.1.5, covering the 
testing of Suppression Chamber-Drywell Vacuum Breakers and the Drywell-
to-Suppression Chamber Bypass Leakage Test.
    Date of issuance: November 7, 2001
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 149 and 135
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38761).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments are contained 
in a Safety Evaluation dated November 7, 2001.
    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: July 9, 2001
    Brief description of amendments: These amendments revised the 
current Technical Specifications of Limerick Generating Station, Units 
1 and 2, to make them more consistent with changes to Title 10 of the 
Code of Federal Regulations, Section 50.59.
    Date of issuance: As of date of issuance and shall be implemented 
within 60 days.
    Effective date: November 1, 2001
    Amendment Nos.: 154 and 118
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44170).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 1, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
County, Pennsylvania

    Date of application for amendments: July 9, 2001
    Brief description of amendments: These amendments replaced the term 
``unreviewed safety question'' with ``requires NRC approval pursuant to 
10 CFR 50.59'' in order to provide consistency with the changes to 10 
CFR 50.59, ``Changes, tests, and experiments,'' which became effective 
on March 13, 2001.
    Date of issuance: November 6, 2001
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendments Nos.: 242 and 246.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications and the License.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44170).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 6, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: April 1, 2001, as supplemented 
July 20, 2001.
    Brief description of amendment: This amendment introduces new 
Technical Specification 6.17, ``Technical Specification (TS) Bases 
Control Program'' to provide consistency with the changes to 10 CFR 
50.59 as published in the Federal Register (Volume 64, Number 191) 
dated October 4, 1999.
    Date of issuance: November 15, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 249.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29356).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 15, 2001.
    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: August 22, 2001.
    Brief description of amendments: Revised Technical Specifications 
Section 6.0, ``Administrative Controls,'' to change the title of the 
corporate executive responsible for plant nuclear safety from 
``President-Nuclear Division'' to ``Chief Nuclear Officer.''
    Date of Issuance: November 13, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 178 and 121.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.

[[Page 59517]]

    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50469).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 13, 2001.
    No significant hazards consideration comments received: No.

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 application for amendments: June 12, 2000, as supplemented 
by letters dated November 7, 2000, June 19 and August 17, 2001.
    Brief description of amendments: The amendments would use the 
methodology and the alternative source term (AST) in 10 CFR 50.67 and 
described in NUREG-1465, ``Accident Source Terms for Light-Water 
Nuclear Power Plants,'' and Regulatory Guide 1081, ``Alternative 
Radiological Source Terms for Evaluating the Radiological Consequences 
of Design-Basis Accidents at Boiling and Pressurized Water Reactors.'' 
Implementing the AST of 10 CFR 50.67 results in a new acceptance 
criterion for 10 CFR Part 50, Appendix A, General Design Criterion 19, 
of 5 rem total effective dose equivalent. The licensee determined that 
use of the revised analysis assumptions, methodology, and acceptance 
criterion required prior Nuclear Regulatory Commission (NRC) approval. 
In addition, the NRC requires in 10 CFR 50.67, a license amendment to 
implement the AST as a replacement for the Technical Information 
Document 14844 source term.
    Date of issuance: November 13, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 258 and 241.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
approve changes to the updated final safety analysis report.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51356).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 13, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 15, 2001.
    Brief description of amendment: The amendment consists of deletion 
of Operating License Condition 2.D, and revision to the Technical 
Specifications (TSs) to remove depiction of railroad tracks in TS 
Figure 4.1-1.
    Date of issuance: November 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 190
    Facility Operating License No. DPR-46: Amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34285).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 16, 2001.
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New York

    Date of application for amendments: February 1, 2001; as 
supplemented on March 1, March 16, March 29, April 5, April 27, May 30, 
June 7, September 10, September 26, September 28, and November 2, 2001.
    Brief description of amendments: The amendments changed the 
operating licenses and associated documents to reflect the transfer of 
Niagara Mohawk Power Corporation's (NMPC's) ownership interest in Nine 
Mile Point Nuclear Station, Unit No. 1, the transfer of the ownership 
interests of NMPC, New York State Electric and Gas Corporation, 
Rochester Gas and Electric Corporation, and Central Hudson Gas & 
Electric Corporation in Nine Mile Point Nuclear Station, Unit No. 2, 
and the transfer of NMPC's operating authority for both units, to Nine 
Mile Point Nuclear Station, LLC. The amendments and corresponding 
license transfers were approved by the U.S. Nuclear Regulatory 
Commission by Order dated June 22, 2001, and Supplemental Order dated 
October 30, 2001.
    Date of issuance: November 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 172 (for Unit 1), 100 (for Unit 2).
    Facility Operating License Nos. DPR-63 and NPF-69: Amendments 
revised the operating licenses (both units), Technical Specifications 
(both units) and Environmental Protection Plan (Unit 2).
    Date of initial notice in Federal Register: April 2, 2001 (66 FR 
17584).
    The staff's related evaluation of the amendments is contained in 
two Safety Evaluations dated June 22 and October 30, 2001.
    No significant hazards consideration comments received: Not 
applicable.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: August 16, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications by deleting Section 5.5.3, ``Post Accident 
Sampling,'' and thereby eliminating the requirements to have and 
maintain the post-accident sampling program. The amendments also 
revised Section 5.5.2, ``Primary Containment Sources Outside 
Containment,'' to reflect the elimination of requirements to maintain 
the post accident sampling system.
    Date of issuance: November 13, 2001.
    Effective date: As of the date of issuance and shall be implemented 
on or before June 28, 2002.
    Amendment Nos.: 123 and 101.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50472).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 13, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 2, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications by deleting Section 6.8.3.d, ``Post Accident 
Sampling,'' and thereby eliminate the requirements to have and maintain 
the post-accident sampling program. The amendments also revise Section 
6.8.3.a, ``Primary Containment Sources Outside Containment,'' to 
reflect the elimination of requirements to maintain the post accident 
sampling system.
    Date of issuance: November 7, 2001.

[[Page 59518]]

    Effective date: As of the date of issuance and shall be implemented 
within 6 months of the date of issuance.
    Amendment Nos.: Unit 1--133; Unit 2--122.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated November 7, 2001.
    No significant hazards consideration comments received: No.

    Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.


    Dated at Rockville, Maryland, this 20th day of November 2001.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 01-29446 Filed 11-27-01; 8:45 am]
BILLING CODE 7590-01-P