[Federal Register Volume 66, Number 172 (Wednesday, September 5, 2001)]
[Notices]
[Pages 46473-46485]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-22137]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section

[[Page 46474]]

189 of the Act. This provision grants the Commission the authority to 
issue and make immediately effective any amendment to an operating 
license upon a determination by the Commission that such amendment 
involves no significant hazards consideration, notwithstanding the 
pendency before the Commission of a request for a hearing from any 
person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 13, 2001 through August 24, 2001. The 
last biweekly notice was published on August 22, 2001 (66 FR 44161).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By October 5, 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 Commission's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). 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.

[[Page 46475]]

    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 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].

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

    Date of amendments request: July 26, 2001.
    Description of amendments request: The proposed amendment modifies 
Administrative Controls Technical Specifications (TSs) 5.5.14.b and 
5.5.14.b.2 such that they are consistent with Title 10 of the Code of 
Federal Regulations (10 CFR), Sec. 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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change replaces the word ``involve'' with 
``require'' and deletes reference to the term ``unreviewed safety 
question'' consistent with 10 CFR 50.59. Deletion of the term 
``unreviewed safety question'' was approved by the Nuclear 
Regulatory Commission with the revision to 10 CFR 50.59. 
Consequently, the probability of an accident previously evaluated is 
not significantly increased. Changes to the TS Bases are still 
subject to 10 CFR 50.59. As a result, the consequences of any 
accident previously evaluated are not significantly affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing plant operation. These changes are 
considered administrative changes and do not modify, add, delete, or 
relocate any technical requirements in the TS.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The proposed changes will not reduce the margin of safety 
because they have no effect on any safety analyses assumptions. 
Changes to the TS Bases are still subject to 10 CFR 50.59, including 
prior Nuclear Regulatory Commission approval if the criteria in 10 
CFR 50.59(c)(2) are met. The proposed changes to TS 5.5.14 are 
considered administrative in nature based on the revision to 10 CFR 
50.59.
    Therefore, this proposed modification 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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Peter Tam (Acting).

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

    Date of amendments request: July 27, 2001.
    Description of amendments request: The proposed amendment modifies 
the conditions and required actions for the control room emergency 
ventilation system (CREVS) of Technical Specification (TS) 3.7.8 for 
Calvert Cliffs Nuclear Power Plant, Units Nos. 1 & 2. Note 2 is being 
added to TS 3.7.8 to specify CREVS train operability requirements 
during the movement of irradiated fuel assemblies. Associated Limiting 
Conditions for Operation (LCO) Action Statements F and G are also being 
modified to be consistent with the addition of Note 2.
    The proposed amendment also modifies the conditions and required 
actions for the control room emergency temperature system (CRETS) of TS 
3.7.9. The existing note in TS 3.7.9 is being modified to specify CRETS 
train operability requirements during the movement of irradiated fuel 
assemblies. LCO Action Statements C and D are also being modified to be 
consistent with the addition of Note 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed changes will modify the conditions and required 
actions for the Control Room Emergency Ventilation System (CREVS) 
and the Control Room Emergency Temperature System (CRETS) to reflect 
the licensing basis for movement of irradiated fuel assemblies. The 
CREVS and CRETS mitigate the consequences of an accident and do not 
initiate an accident. The CREVS provides protection to the control 
room operators in the event of a radioactive release. The CRETS 
provides protection to the Control Room by maintaining the 
temperature below the required limit. Therefore, changing the 
Conditions, Required Actions, and Completion Times for the CREVS and 
CRETS does not increase the probability of an accident.
    As described in the Updated Final Safety Analysis Report 
(UFSAR), the CREVS and CRETS mitigate the consequences of six 
accidents. All but the fuel handling accident are postulated to 
occur during Modes 1, 2, 3, or 4. The fuel handling accident is only

[[Page 46476]]

postulated to occur during the movement of irradiated fuel 
assemblies. The changes proposed would only alter the response to 
the loss of one CREVS or CRETS train during the movement of 
irradiated fuel assemblies. Since a single failure is not required 
to be postulated during the response to a fuel handling accident, 
having one CREVS or CRETS train out-of-service during fuel movement 
would not result in a change to the ability of the CREVS or CRETS to 
mitigate the consequences of a design basis fuel handling accident. 
The loss of one CREVS or CRETS train during Modes 1, 2, 3, or 4 is 
covered by other Conditions, and those Conditions have not been 
changed by this request. Therefore, the ability of the CREVS or 
CRETS to respond to any design basis accident would not be 
diminished by this proposed change.
    Therefore, the proposed Technical Specification changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed changes do not involve a change in the operation of 
the plant and no new accident initiation mechanism is created by the 
proposed changes. The operations of the CREVS or CRETS are not 
altered by the proposed changes. The proposed changes do not change 
the licensing basis requirements for the CREVS or CRETS response to 
the accidents described in the UFSAR. No plant changes will be made 
as a result of this request. No conditions have been created by this 
request that might result in a new accident that has not been 
previously analyzed. Therefore, the proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.
    The margin of safety created by the response of the CREVS or 
CRETS to various accidents has not been reduced by the proposed 
changes in the Conditions, Required Actions, or Completion Times. 
These changes merely clarify the Technical Specification so that the 
licensing basis is more accurately reflected. The fuel handling 
accident does not assume a single failure occurs during the plant 
response to the event; therefore, the loss of a single CREVS or 
CRETS train does not place the plant outside of the licensing basis. 
This would be reflected in the proposed changes. The changes do not 
alter the operation or response requirements of the CREVS or CRETS. 
The CREVS and CRETS will continue to respond to accidents as 
designed. Operators will continue to be protected as described in 
the UFSAR. Therefore, the margin of safety is not significantly 
reduced by these proposed changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Peter Tam (Acting).

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

    Date of amendments request: July 27, 2001.
    Description of amendments request: The proposed amendment will add 
additional references to Technical Specification (TS) 5.6.5.b for the 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 & 2. The references 
will allow the use of ZIRLOTM clad fuel rods in the Calvert 
Cliffs' reactor cores.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change allows the use of methods required for the 
implementation of ZIRLOTM clad fuel rods in Calvert 
Cliffs Unit Nos. 1 and 2 and the use of current versions of the ECCS 
[emergency core cooling system] performance evaluation models for 
large and small break LOCAs [loss-of-coolant accidents]. The use of 
updated analysis methodologies will not increase the probability of 
an accident because the plant systems will not be operated outside 
of design limits, no different equipment will be operated, and 
system interfaces will not change.
    With ZIRLOTM material introduced in the reactor, 
cores will exist in which ZIRLOTM and Zircaloy-4 clad 
fuel rods are co-resident. Fuel rods clad with each material will be 
evaluated based on the approved topical report.
    The use of the three additional methodologies will not increase 
the consequences of an accident because Limiting Conditions for 
Operation (LCOs) will continue to restrict operation to within the 
regions that provide acceptable results, and Reactor Protective 
System (RPS) trip setpoints will restrict plant transients so that 
the consequences of accidents will be acceptable. Also, the 
consequences of the accidents will be calculated using NRC accepted 
methodologies.
    The cores that will exist with ZIRLOTM and/or 
Zircaloy-4 clad fuel in the reactor will not increase the 
consequences of an accident. Operation within the LCOs and RPS 
setpoints will continue to restrict plant transients so that the 
consequences of accidents will be acceptable.
    Therefore, the proposed Technical Specification changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with any existing equipment, alter the equipment's 
function or change the method of operating the equipment. The 
proposed change does not alter plant conditions in a manner that 
could affect other plant components. The proposed change does not 
cause any existing equipment to become an accident initiator. The 
ZIRLOTM clad fuel rod design does not introduce features 
that could initiate an accident. Therefore, the proposed change does 
not create the possibility of a new or different [kind] of accident 
from any accident previously evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.
    Safety Limits ensure that Specified Acceptable Fuel Design 
Limits are not exceeded during steady state operation, normal 
operational transients, and anticipated operational occurrences. All 
fuel limits and design criteria shall be met based on the approved 
methodologies defined in the topical reports. The RPS in combination 
with the LCOs, will continue to prevent any anticipated combination 
of transient conditions for reactor coolant system temperature, 
pressure and thermal power level that would result in a violation of 
the Safety Limits. Therefore, the proposed changes will not involve 
a significant reduction in the margin of safety.
    The safety analyses determine the LCO settings and RPS setpoints 
that establish the initial conditions and trip setpoints, which 
ensure that the Design Basis Events (Postulated Accidents and 
Anticipated Operational Occurrences) analyzed in the Updated Final 
Safety Analysis Report produce acceptable results. Also all fuel 
limits and design criteria shall be satisfied. The Design Basis 
Events that are impacted by the implementation of ZIRLOTM 
cladding will be analyzed using the NRC accepted methodology 
described in CENPD-404-P.
    The change in the fuel rod cladding material and the use of the 
current ECCS performance evaluation models will not involve a 
reduction in the margin of safety because acceptable results for the 
impacted Design Basis Events will be maintained.
    Therefore, the margin of safety is not significantly reduced by 
this proposed change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

[[Page 46477]]

    NRC Section Chief: Peter Tam, Acting.

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

    Date of amendments request: August 1, 2001.
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TS) to support a full-scope 
application of an Alternative Source Term (AST). The AST analyses were 
performed following the guidance in Regulatory Guide 1.183, 
``Alternative Radiological Source Terms For Evaluating Design Basis 
Accidents At Nuclear Power Reactors,'' dated July 2000, and Standard 
Review Plan Section 15.0.1, ``Radiological Consequences Analyses Using 
Alternative Source Terms.'' 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 BSEP systems affected by implementation of the Alternative 
Source Term analyses and the relaxations associated with TSTF 
[Technical Specification Task Force]-51, Revision 2, are not 
initiators of any design basis accidents. Therefore, because design 
bases accident initiators are not being altered by adoption of the 
Alternative Source Term analyses and the relaxations associated with 
TSTF-51, Revision 2, the probability of an accident previously 
evaluated is not affected. The Alternative Source Term does not 
affect the design or normal operation of the facility. Rather, once 
the occurrence of the accident has been postulated, the Alternative 
Source Term is an input used to evaluate the consequences of an 
accident. Implementation of the Alternative Source Term has been 
evaluated for the limiting design basis accidents at BSEP, and it 
has been demonstrated that the dose consequences of those limiting 
design bases accidents are within the regulatory guidance provided 
by the NRC in Regulatory Guide 1.183 and Standard Review Plan 
Section 15.0.1. For a fuel handling accident, the AST analyses 
demonstrate acceptable doses, within regulatory limits, without 
credit for secondary containment or automatic isolation of the 
Control Room. As such, the consequences of an accident previously 
evaluated are not affected.
    Based on the above, the proposed license amendments do not 
involve an 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.
    The BSEP systems affected by implementing the Alternative Source 
Term changes and the changes associated with TSTF-51, Revision 2, do 
not alter any design bases accident initiators or create new types 
of accident precursors. In addition, these changes do not affect the 
design function or mode of operation of systems, structures, or 
components in the facility such that new equipment failure modes are 
created. Therefore, the proposed license amendments will not create 
the possibility of 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 changes proposed are associated with the implementation of a 
new licensing basis for BSEP. Approval of the change from the 
original source term, developed in accordance with TID-14844, to a 
new Alternative Source Term, as described in NUREG-1465, ``Accident 
Source Terms for Light-Water Nuclear Power Plants, Final Report,'' 
dated February 1, 1995, is being requested. The results of the 
accident analyses, revised in support of the proposed license 
amendments, are subject to revised acceptance criteria. These 
analyses have been performed using conservative methodologies, as 
specified in Regulatory Guide 1.183. Safety margins have been 
evaluated and margin has been retained to ensure that the analyses 
adequately bound the postulated limiting event scenarios. The dose 
consequences of these limiting events are within the acceptance 
criteria presented in 10 CFR 50.67, ``Alternative source term,'' and 
Regulatory Guide 1.183.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries, as well as the 
Control Room and Emergency Operations Facility/Technical Support 
Center, are within corresponding regulatory limits. Specifically, 
the margin of safety for these accidents is considered to be that 
provided by meeting the applicable regulatory limits, which for 
three of five event scenarios (i.e., the control rod drop accident, 
the fuel handling accident, and one of the two limits for a main 
steam line break accident), is conservatively set below the 10 CFR 
50.67 limit. With respect to the Control Room personnel doses, the 
margin of safety is the difference between the 10 CFR 50.67 limits 
and the regulatory limit defined by 10 CFR 50, Appendix A, General 
Design Criterion 19.
    Since the proposed changes continue to ensure that the doses at 
the exclusion area and low population zone boundaries, as well as 
the Control Room are within corresponding regulatory limits, the 
proposed license amendments do not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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.

Consolidated Edison Company of New York, Inc., Docket No. 50-003, 
Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York

    Date of amendment request: July 13, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 6.12, ``High Radiation Area,'' to 
delete the administrative requirements for the control of access to 
high radiation areas. The control of access to these areas is assured 
by plant radiation protection programs that comply with 10 CFR 20.1601 
requirements by using the alternate method in Regulatory Guide 8.38, 
``Control of Access to High and Very High Radiation Areas of Nuclear 
Power Plants,'' June 1993.
    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 in the consequences of an accident 
previously evaluated?
    The proposed TS change is administrative in nature. It involves 
deleting specific requirements for complying with a subparagraph of 
10 CFR [part] 20 for the purpose of controlling access to high 
radiation areas. Accident evaluations do not consider the effects of 
methods of controlling access to high radiation areas. The proposed 
changes do not result in a change to the design or operation of [* * 
*] any plant structure, system, or component. Therefore any 
assumptions of the operability or performance of any structure, 
system, or component in accident evaluations are unchanged.
    Therefore, there is no increase in the probability or in the 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed change is administrative in nature. The methods of 
controlling access to high radiation areas do not affect the design 
or operation of any plant structure, system, or component. 
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 proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed TS change is administrative in nature. It involves 
deleting specific

[[Page 46478]]

requirements for complying with a subparagraph of 10 CFR [part] 20. 
However, effective compliance with 10 CFR [part] 20 is mandated by 
[* * *] another IP1 [Indian Point Nuclear Generating Station, Unit 
1] TS provision. The effectiveness of Con Edison [Consolidated 
Edison Company of New York, Inc.] compliance with 10 CFR [part] 20 
is not adversely affected by this change. In addition, this change 
does not affect any design function for or the operation of any 
plant structure, system, or component.
    Therefore, the change does not affect any of the safety analyses 
or any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Robert A. Gramm.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: August 14, 2001.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Surveillance Requirement 3.3.5.2 by 
changing the Engineered Safeguards Protective System Analog Instrument 
channel functional test frequency from 31 days to 92 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:

    Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the 
determination that this amendment request involves a No Significant 
Hazards Consideration by applying the standards established by the 
NRC regulations in 10 CFR 50.92. This ensures 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:
    No. This is a proposed change to the Technical Specification 
(TS) 3.3.5 Engineered Safeguards Protective System (ESPS) Analog 
Instrumentation, Surveillance Requirement (SR) 3.3.5.2 for the 
channel functional test. The proposed change to TS 3.3.5 ESPS Analog 
Instrumentation, SR 3.3.5.2 will extend the current 31 day 
surveillance frequency to a 92 day surveillance frequency. The 
proposed change does not alter the method of operating or 
configuration for any Structure, System, or Component.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    No. The ESPS Analog Instrumentation provides the necessary 
actuation of the Engineered Safety Features based on the Reactor 
Coolant and/or Reactor Building pressure. The proposed revision to 
the frequency for SR 3.3.5.2 will not alter the actuation of the 
Engineered Safety Features. The channel functional testing of the 
ESPS Analog Instrumentation will continue to be performed in an 
acceptable timeframe following the implementation of the proposed 
change.
    (3) Involve a significant reduction in a margin of safety.
    No. The proposed revision to the frequency for SR 3.3.5.2 will 
not impact the operation of the ESPS Analog Instrumentation. In 
addition, the channel functional testing of the ESPS Analog 
Instrumentation will continue to be performed in an acceptable 
timeframe following the implementation of the proposed change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: August 23, 2001.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to eliminate the requirement to 
move control element assembly #43 for the remainder of Cycle 15.
    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 operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    One function of the CEAs [control element assemblies] is to 
provide a means of rapid negative reactivity addition into the core. 
This occurs upon receipt of a signal from the Reactor Protection 
System. This function will continue to be accomplished with the 
approval of the proposed change. Typically, once per 92 days each 
CEA is moved at least five inches to prove operability. Operability 
of a CEA requires the CEA be trippable and free from mechanical 
binding, i.e., moveable. CEA #43 is operable. However, due to 
abnormal coil voltage on two of the five coils that move CEA #43, if 
CEA #43 were moved to perform the SR [surveillance requirement], it 
is possible that a drop rod incident could occur. The misoperation 
of a CEA, which includes a drop rod incident, is an abnormal 
occurrence and has been evaluated as part of the ANO-2 [Arkansas 
Nuclear One, Unit 2] accident analysis. The proposed change would 
eliminate the requirement to move CEA #43 every 92 days and 
therefore eliminate the potential of CEA misoperation, associated 
down power, and challenge to the plant.
    If a reactor trip signal were generated, CEA #43 has been 
demonstrated to be operable and will drop into the core along with 
the remaining CEAs to ensure reactor shutdown. No modifications are 
proposed to the Reactor Protection System or associated Control 
Element Drive Mechanism Control System logic. The accident 
mitigation features of the plant are not affected by the proposed 
amendment.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will 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?
    CEA #43 is operable, both moveable and trippable. The proposed 
change will not introduce any new design changes or systems. If a 
reactor trip were generated, CEA #43 will drop into the core along 
with the remaining CEAs to ensure reactor shutdown. The proposed 
change does not establish a potential for a new accident precursor.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    CEA #43 will continue to have the same capability to mitigate an 
accident as it had prior to approval of the proposed TS [technical 
specification] change. CEA #43 is moveable and trippable.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Section Chief: Robert A. Gramm.

[[Page 46479]]

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

    Date of amendment request: July 18, 2001.
    Description of amendment request: The submittal requests a change 
to Technical Specifications (TS) Definitions 1.12 and 1.25, the effect 
of which will be to allow either an allocated or a measured response 
time to be utilized for the sensors in the Reactor Protective System 
and Engineered Safety Features Actuation System instrument loops.
    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 operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed amendment to Technical Specifications (TS) Definitions 
1.12 and 1.25 allows substitution of an allocated sensor response 
time in lieu of measuring sensor response time. Response time 
testing is not an initiator of any accident previously evaluated. 
Further, overall system response time will continue to meet 
Technical Specification requirements. The allocated sensor response 
times allowed in lieu of measurement have been determined to 
adequately represent the response time of the components such that 
the safety systems utilizing those components will continue to 
perform their accident mitigation function as assumed in the safety 
analysis.
    Therefore, this change does not involve a significant increase 
in the probability of consequences of any accident previously 
evaluated.
    2. Will 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?
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed amendment to TS Definitions 1.12 and 1.25 
allows the substitution of an allocated sensor response time in lieu 
of sensor response time testing for selected components. The 
proposed change does not involve a physical alteration of the plant 
(no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The use of 
allocated response times in lieu of measured response times result 
in no physical change to the plant.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The proposed amendment does not involve a significant reduction 
in a margin of safety. The proposed amendment to TS 1.1, 
Definitions, allows the substitution of an allocated sensor response 
time in lieu of measured sensor response time for certain pressure 
sensors. The allocated pressure sensor response times allowed in 
lieu of measurement have been determined to adequately represent the 
response time of the components such that the safety systems 
utilizing those components will continue to perform their accident 
mitigation function as assumed in the safety analysis.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: N. S. Reynolds, Esquire, Winston and Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. 50-237, Dresden Nuclear 
Power Station, Unit 2, Grundy County, Illinois

    Date of amendment request: June 6, 2001.
    Description of amendment request: The proposed amendment would 
revise the values of the Safety Limit for the Minimum Critical Power 
Ratio in Technical Specification 2.1.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC 
approved methods to ensure that fuel performance during normal, 
transient and accident conditions is acceptable. The proposed change 
conservatively establishes the safety limit for the minimum critical 
power ratio (SLMCPR) for Dresden Nuclear Power Station (DNPS), Unit 
2, Cycle 18 such that the fuel is protected during normal operation 
and during any plant transients or anticipated operational 
occurrences.
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operations. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during any transients or 
anticipated operational occurrences. Operational limits will be 
established based on the proposed SLMCPR to ensure that the SLMCPR 
is not violated during all modes of operation. This will ensure that 
the fuel design safety criteria (i.e., that at least 99.9 percent of 
the fuel rods do not experience transition boiling during normal 
operation and anticipated operational occurrences) is met. Since the 
operability of plant systems designed to mitigate any consequences 
of accidents has not changed, the consequences of an accident 
previously evaluated are not expected to increase.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation. The proposed change does not involve 
any modifications of the plant configuration or allowable modes of 
operation. The proposed change to the SLMCPR assures that safety 
criteria are maintained for DNPS, Unit 2, Cycle 18.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The value of the proposed SLMCPR provides a margin of safety by 
ensuring that no more than 0.1 percent of the rods are expected to 
be in boiling transition if the [minimum critical power ratio] MCPR 
limit is not violated. The proposed change will ensure the 
appropriate level of fuel protection. Additionally, operational 
limits will be established based on the proposed SLMCPR to ensure 
that the SLMCPR is not violated during all modes of operation. This 
will ensure that the fuel design safety criteria (i.e., that at 
least 99.9 percent of the fuel rods do not experience transition 
boiling during normal operation as well as anticipated operational 
occurrences) are met.
    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

[[Page 46480]]

satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment 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.

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

    Date of amendment request: May 30, 2001.
    Description of amendment request: The proposed amendment would 
change the Technical Specification (TS) 5.5.7.a, b, and c, to update 
the Ventilation Filter Testing Program at Cooper Nuclear Station.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The District has evaluated each of the proposed TS changes in 
accordance with the criteria set forth in 10 CFR 50.92 and has 
determined that the proposed changes do not involve a significant 
hazards consideration.
    The determination that the proposed changes do not involve a 
significant hazards consideration is based on an evaluation of these 
changes against each of the criteria in 10 CFR 50.92. The criteria 
and the conclusions of the evaluation are presented below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The application of the 1989 version of ASME N510 will not change 
any of the surveillance requirements for operability of the SGT or 
the CREF. The changes with respect to RG 1.52 are editorial in 
nature and will not result in any changes in surveillance 
requirements. Since SGT and CREF are ESF systems and not accident 
initiators the probability of an accident evaluated in the Updated 
Safety Analysis Report will not be increased. As such, the 
probability of occurrence for a previously analyzed accident is not 
significantly increased.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed for the analysis, the availability 
and successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. This change does not affect the performance 
of any credited equipment. These details of testing are not analysis 
assumptions. Based on this evaluation, there is no significant 
increase in the consequences of a previously analyzed event.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
is no change being made to the parameters within which the plant is 
operated. There are no setpoints, at which protective or mitigative 
actions are initiated, affected by this change. This change will not 
alter the manner in which equipment operation is initiated, nor will 
the function demands on credited equipment be changed. The change 
does not result in alteration of the procedures which ensure the 
plant remains within analyzed limits, and no change is being made to 
the procedures relied upon to respond to an off-normal event. As 
such, no new failure modes are being introduced. The change does not 
alter assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. Sufficient equipment remains available to actuate 
upon demand for the purpose of mitigating an analyzed event. The 
proposed change, which replaces references to ASME N510-1980 with 
references to ASME N510-1989, is acceptable because the tests 
continue to require appropriated confirmation of the assumed 
function of the systems (and thereby assure continued operability), 
and more accurately presents acceptable testing conditions. The 
changes with respect to RG 1.52 are editorial in nature and do not 
change existing surveillances. There is no detrimental impact on any 
equipment design parameter, and the plant will still be required to 
operate within prescribed limits. Therefore, the change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499
    NRC Section Chief: Robert A. Gramm.

PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric 
Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: May 31, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 2.1.1.2 to reflect the Unit 1 Cycle 13 ( 
U1C13) minimum critical power ratio (MCPR) safety limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    No. The proposed change to the MCPR Safety Limit does not 
directly or indirectly affect any plant system, equipment, 
component, or change the way in which the plant is operated. Thus, 
this proposed amendment does not involve a significant increase in 
the probability of occurrence of an accident previously evaluated.
    Prior to the startup of U1C13, licensing analyses are performed 
(using NRC approved methodology referenced in Technical 
Specification Section 5.6.5.b) to determine changes in the critical 
power ratio as a result of anticipated operational occurrences. 
These results are added to the MCPR Safety Limit values proposed 
herein to generate the MCPR operating limits in the U1C13 COLR [Core 
Operating Limits Report]. These limits could be different from those 
specified for the U1C12 COLR. The COLR operating limits thus assure 
that the MCPR Safety Limit will not be exceeded during normal 
operation or anticipated operational occurrences.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously analyzed?
    No. The change to the MCPR Safety Limits and the U1C13 core 
loading which it supports does not directly or indirectly affect any 
plant system, equipment, or component (other than the core itself) 
and therefore does not affect the failure modes of any of these. 
Thus, the proposed changes do not create the possibility of a 
previously unevaluated operator error or a new single failure.
    Therefore, this proposed amendment does not involve a 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    No. Since the proposed changes do not affect any plant system, 
equipment, or component, the proposed change will not jeopardize or 
degrade the function or operation of any plant system or component 
governed by Technical Specifications. The proposed MCPR Safety 
Limits do not involve a significant reduction in the margin of 
safety as currently defined in the Bases of the applicable Technical 
Specification sections, because the MCPR Safety Limits calculated 
for U1C13 preserve the required margin of safety.

[[Page 46481]]

    Therefore these changes do not involve a significant reduction 
in [a] margin of safety.

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

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

    Date of amendment request: June 1, 2001, as supplemented June 13, 
2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.7.1, Residual Heat 
Removal Service Water (RHRSW) System and Ultimate Heat Sink (UHS), to 
address previously unidentified single failure vulnerabilities when one 
or more RHRSW subsystems are inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Implementation of the subject changes reduces the probability of 
occurrence and the probability of adverse consequences of accidents 
previously evaluated. Inclusion of the large array valves and the 
bypass valves to the Technical Specifications (TS) recognizes their 
importance to safe shutdown. The administrative controls that TS's 
invoke increases the probability that potential inoperability of 
these valves is controlled and managed in a manner commensurate with 
their risk significance.
    Reducing the completion time for RHRSW subsystem inoperable 
conditions recognizes their importance to safe shutdown commensurate 
with their risk significance.
    These changes do not affect the design or operation of the 
affected components/systems and serves to increase the level of 
administrative control for the UHS and RHRSW system that will help 
to ensure the ability to achieve safe shutdown.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously analyzed?
    The subject changes apply Technical Specification administrative 
controls to the UHS bypass and large array valves and shortens the 
completion times applicable to RHRSW inoperable conditions. The 
design and operation of the affected components and systems is not 
affected.
    Application of these administrative controls does not involve a 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    Implementation of the subject changes increases the margin of 
safety since these changes add Technical Specification controls to 
components not currently addressed in the Technical Specifications 
and reduces the completion times for subsystems currently addressed 
in the Technical Specifications. These changes better account for 
the affected components/systems impact on safe shutdown.
    Therefore these changes do not involve a significant reduction 
in [a] margin of safety.

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

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: May 3, 2001.
    Description of amendment request: The licensee proposed to amend 
Ginna Station Improved Technical Specifications (ITS) to reflect the 
design changes to the actuation circuitry associated with the Control 
Room Emergency Air Treatment System (CREATS). The proposed design 
changes consist of replacing the current diverse radiation monitors 
with two Gieger-Mueller (GM) tubes powered from two separate safety-
related power supplies which are configured into two redundant 
actuation logic trains, including manual initiation. The design changes 
is intended to increase system reliability by providing redundancy and 
reducing spurious actuations. As a result of the proposed design 
changes, the licensee requested that the following changes be made to 
the Limiting Condition for Operation (LCO) 3.3.6 for the CREATS 
Actuation Instrumentation:
    a. Add a new Condition to require immediately placing the CREATS in 
the emergency mode of operation upon the loss of two instrument 
channels/trains.
    b. Add a new surveillance requirement involving a CHANNEL CHECK of 
the Control Room Radiation Intake Monitors.
    c. Revise Table 3.3.6-1 to increase the number of trains of Manual 
Initiation, and Automatic Actuation Logic and Actuation Relays, from 
one train to two trains.
    d. Extend the Completion Time of the Required Action for a loss of 
one channel/train from 1 hour to 7 days as the result of installing 
redundant channels/trains.
    e. Revise Table 3.3.6-1 to remove reference to the Iodine, Noble 
Gas, and Particulate Control Room Radiation Intake Monitors. These 
monitors will be replaced by the two new GM tubes.
    f. Revise Table 3.3.6-1 to replace the column heading ``Trip 
Setpoint'' with ``Allowable Value.''
    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 reviewed the licensee's analysis against 
the three standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed ITS changes listed above will not increase the 
probability of an accident previously evaluated because the CREATS 
actuation system is not an accident initiator as this system performs 
only mitigative functions. In particular, the system is designed to 
provide a protective environment from which the operators can control 
the plant following an uncontrolled release of radioactivity during a 
design-basis accident. The proposed design changes (increase system 
redundancy and reliability) and the ITS changes associated with LCO 
3.3.6 (i.e., action statements for loss of instrument channels/trains, 
channel check requirements, etc.,) will only ensure that the CREATS 
will continue to perform its safety functions and that the consequences 
of an accident previously evaluated will not increase.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed ITS changes listed above will not create the 
possibility of a new or different kind of

[[Page 46482]]

accident from any accident previously evaluated because the CREATS 
actuation system is not an accident initiator as this system performs 
only mitigative functions. The proposed change creates no new 
functional interactions with existing plant equipment nor does it 
introduce any new failure mode or mechanisms which could lead to 
reactor core damage or fission product release.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. The proposed ITS changes will not 
adversely affect the performance of the CREATS, nor will they affect 
the ability of the system to perform their intended functions. The 
reason being that the proposed amendment does not involve any new 
acceptance criteria, analytical limits, or evaluation models which 
could affect operator dose limits.
    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: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Section Chief: P. Tam, Acting.

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

    Date of amendment request: April 25, 2001 as supplemented by letter 
dated July 31, 2001.
    Brief description of amendments: The proposed license amendments 
would change the Technical Specifications (TS) to allow a one-time only 
change to TS 3.8.1, ``AC Sources--Operating,'' Action A.3, by extending 
the required Completion Time for restoration of an inoperable offsite 
circuit from 72 hours to 21 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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed one time Technical Specification Completion Time 
(CT) extension does not significantly increase the probability of 
occurrence of a previously evaluated accident because the startup 
transformer [ST] XST2 is not an initiator of previously evaluated 
accidents involving a loss of offsite power. The proposed changes to 
the Technical Specification CT do not affect any of the assumptions 
used in the deterministic or the Probabilistic Safety Assessment 
(PSA) analysis relative to loss of offsite power initiating event 
frequency.
    The proposed one time Technical Specification CT extension will 
continue to provide assurance that the sources of power to 6.9 kV AC 
[kilovolt alternating current] buses perform their function when 
called upon. Extending the Technical Specification CT to 21 days 
does not affect the design of XST2, the operational characteristics 
of XST2, the interfaces between XST2 and other plant systems, the 
function, or the reliability of XST2. Thus, 6.9 kV AC components 
will be capable of performing either accident mitigation function 
and there is no impact to the radiological consequences of any 
accident analysis.
    To fully evaluate the effect of the proposed change, PSA methods 
and deterministic analysis were utilized. The results of this 
analysis show no significant increase in the Core Damage Frequency.
    The Maintenance Rule (a)(4) risk management program assesses 
risk based on plant status. It requires the consideration of other 
measures to mitigate consequences of an accident occurring while a 
ST is unavailable.
    The proposed changes do not alter the operation of any plant 
equipment assumed to function in response to an analyzed event or 
otherwise increase its failure probability. Therefore, these changes 
do not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    These proposed changes do not change the design, configuration, 
or method of operation of the plant. The proposed activity involves 
a change to the allowed plant mode for the performance of preventive 
maintenance that will ensure the inherent reliability of the XST2 
Startup Transformer is maintained. No physical or operational change 
to the ST or supporting systems are made by this activity. Since the 
proposed change does not involve a change to the plant design or 
operation, no new system interactions are created by this change. 
The proposed Technical Specification change does not produce any 
parameters or conditions that could contribute to the initiation of 
accidents different from those already evaluated in the Final Safety 
Analysis Report.
    The proposed change only addresses the time allowed to restore 
the operability of XST2. Thus the proposed Technical Specification 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the Limiting Conditions for 
Operation or their Bases that are used in the deterministic analysis 
to establish any margin of safety. PSA evaluations were used to 
evaluate the proposed change, and these evaluations determined that 
the net changes are either risk neutral or risk beneficial. The 
proposed activity involves a one time change to Allowed Outage 
Times.
    The proposed change does not involve a change to the plant 
design or operation and thus does not affect the design of the ST, 
the operation characteristics of the ST, the interfaces between the 
ST and other plant systems, or the function or reliability of the 
ST. Because ST performance and reliability will continue to be 
ensured by the proposed one time Technical Specification change, the 
proposed changes do not result in a reduction in the margin of 
safety.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.NRC Section Chief: 
Robert A. Gramm.

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

    Date of amendment request: August 7, 2001 (ET 01-0021).
    Description of amendment request: The amendment would add the 
following to the Wolf Creek Generating Station (WCGS) Technical 
Specifications (TSs): (1) The phrase, ``or if open, capable of being 
closed'' to Limiting Condition for Operation (LCO) 3.9.4 for the 
equipment hatch, during core alterations or movement of irradiated fuel 
assemblies inside containment, and (2) the requirement to verify the 
capability to install the equipment hatch in a new Surveillance 
Requirement (SR) 3.9.4.2. Nothing is proposed to be deleted from the 
TSs. Existing SR 3.9.4.2 would be renumbered SR 3.9.4.3, but would not 
otherwise be changed. Item (1) will allow the equipment hatch to be 
open during the conditions stated above.
    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

[[Page 46483]]

consequences of an accident previously evaluated.
    The proposed changes will allow the equipment hatch to be open 
during CORE ALTERATIONS and movement of irradiated fuel assemblies 
inside containment. The status of the equipment hatch during 
refueling operations has no affect on the probability of the 
occurrence of any accident previously evaluated. The proposed 
revision does not alter any plant equipment or operating practices 
in such a manner that the probability of an accident is increased. 
Since the consequences of a fuel handling accident inside 
containment with an open equipment hatch are bounded by the current 
analysis described in the USAR [Updated Safety Analysis Report for 
WCGS] and the probability of an accident is not affected by the 
status of the equipment hatch, the proposed change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not create any new failure modes for any 
system or component, nor do they adversely affect plant operation. 
No new equipment will be added and no new limiting single failures 
will be created. The plant will continue to be operated within the 
envelope of the existing safety analysis.
    Therefore, the proposed changes do not create 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 previously determined radiological dose consequences for a 
fuel handling accident inside containment with the air lock doors 
open remain bounding for the proposed changes. These previously 
determined dose consequences were determined to be well within the 
limits of 10 CFR 100 and they meet the acceptance criteria of SRP 
[Standard Review Plan, NUREG-0800] section 15.7.4 and GDC [General 
Design Criteria of Appendix A to 10 CFR Part 50] 19.
    Therefore, the proposed changes do 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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licensess

    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].

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

    Date of amendment request: January 24, 2001, as supplemented by 
letters dated July 18 and August 3, 2001.
    Brief description of amendment: The request consists of a change to 
Technical Specification 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs),'' to permit the operation of the Inclined Fuel Transfer System 
(IFTS) bottom valve after removal of the IFTS primary containment 
isolation blind flange while the containment is required to be 
operable.
    Date of issuance: August 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 117.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15921). The supplemental letters dated July 18 and August 3, 2001, 
provided additional information that did not expand the scope of the 
application as originally noticed, and did not change the Nuclear 
Regulatory Commission (the Commission) staff's proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 2001.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2, Pope County, Arkansas

    Date of amendment request: January 27, 2000, as supplemented by 
letters dated March 1, June 12, and July 26, 2001.
    Brief description of amendments: The amendments allow the qualified 
condensate storage tank to be used for both units and defines new 
minimum volume requirements for the tank depending on whether Arkansas 
Nuclear One, Unit 1, Arkansas Nuclear One, Unit 2, or both units are 
aligned to the tank. The total volume requirements, the allowable 
alternative alignment for ANO-2, and other aspects of the Technical 
Specifications (TSs) are unaffected by the change.
    Date of issuance: August 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 214, 232.
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29352).
    The supplemental letters dated June 12 and July 26, 2001, provided 
additional information and revised TSs that did not expand the scope of 
the

[[Page 46484]]

application or change the initial proposed no significant hazards 
consideration determination which addressed the original application 
and supplement dated March 1, 2001.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 16, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 10, 2001.
    Brief description of amendment: The proposed amendment will provide 
an alternative method for complying with the Limiting Conditions for 
Operation (LCO) requirements of Technical Specification 3.3.4.1, ``End 
of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation,'' and 
require that an additional REQUIRED ACTION be added to CONDITION B as 
REQUIRED ACTION B.2.
    Date of issuance: August 10, 2001.
    Effective date: August 10, 2001.
    Amendment No.: 148.
    Facility Operating License No. NPF-29: Amendment revises the TS.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration, are contained in a Safety Evaluation dated 
August 10, 2001.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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 1, 2000.
    Brief description of amendments: Add a Technical Specification (TS) 
section regarding mechanical vacuum pump trip instrumentation.
    Date of issuance: August 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 186 and 181.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 16, 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: April 16, 2001.
    Brief description of amendments: The amendments change the 
reference in Technical Specification 5.5.6, ``Inservice Inspection 
Program for Post Tensioning Tendons,'' from Regulatory Guide 1.35, 
``Inservice Inspection of Ungrouted Tendons in Prestressed Concrete 
Containments,'' Revision 3, 1989, to a reference to Subsection IWL, 
``Requirements of Class CC Concrete Components of Light-Water Cooled 
Power Plants,'' of Section XI, ``Inservice Inspection,'' of the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code, and delete the applicability of Surveillance Requirement 
(SR) 3.0.2 to TS Section 5.5.6. SR 3.0.2 allows the surveillance to be 
performed within 1.25 times the interval specified in the 
surveillance's frequency.
    Date of issuance: August 16, 2001.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 148 and 134.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31707).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 16, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 13, 2001.
    Brief description of amendment: The amendment revises TS 5.3 to 
permit lead-test-assemblies to be used, regardless of clad material, as 
long as the Nuclear Regulatory Commission has generically approved the 
fuel assembly design for use in pressurized water reactors.
    Date of issuance: August 13, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 156.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36342).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 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: December 20, 2000, as supplemented by 
letters dated February 1 and 28, and June 12, 2001.
    Brief description of amendments: The amendments revised the 
technical specification (TS) requirements and authorized revision of 
the Technical Requirements Manual provisions applicable when actions 
direct suspension of operations involving positive reactivity changes. 
The proposed changes remove the requirement not to make positive 
reactivity changes during certain plant conditions, and limit the 
reactivity changes that are allowed to those that will continue to 
assure appropriate reactivity limits are met. Related changes to the 
Bases were also made. In addition, an administrative TS change was made 
to remove a footnote regarding an alternate onsite emergency power 
source, which is no longer applicable.
    Date of issuance: August 13, 2001.
    Effective date: August 13, 2001.
    Amendment Nos.: Unit 1-128; Unit 2-117.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications and authorized revision of the 
Technical Requirements Manual.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9387).
    The February 1 and 28, and June 12, 2001, supplemental letters 
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 amendments is contained 
in a Safety Evaluation dated August 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: December 20, 2000.

[[Page 46485]]

    Brief description of amendments: The amendments delete License 
Condition 2.G, ``Reporting to the Commission,'' and Technical 
Specification 6.6.1.a, ``Reportable Event Action.''
    Date of issuance: August 16, 2001.
    Effective date: The amendments are effective as of the date of 
their issuance.
    Amendment Nos.: Unit 1--129; Unit 2--118.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses and the Technical 
Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31715).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 16, 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: February 28, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) to eliminate periodic response time 
testing requirements on selected sensors and selected protection 
channels, and modified TS Section 1.0 Definitions for ``ENGINEERED 
SAFETY FEATURE (ESF) RESPONSE TIME'' and ``REACTOR TRIP SYSTEM (RTS) 
RESPONSE TIME'' to provide for verification of response time for 
selected components. The associated Bases were also revised.
    Date of issuance: August 21, 2001.
    Effective date: The amendments are effective as of the date of 
their issuance and shall be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: Unit 1--130; Unit 2--119.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31716).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 21, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 28th of August 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-22137 Filed 9-4-01; 8:45 am]
BILLING CODE 7590-01-P