[Federal Register Volume 66, Number 26 (Wednesday, February 7, 2001)]
[Notices]
[Pages 9377-9392]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-3028]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 29, 2001, through February 9, 2001. 
The last biweekly notice was published on January 24, 2001 (66 FR 
7667).

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 an 
opportunity for a

[[Page 9378]]

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's 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 March 9, 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.
    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: Docketing and 
Services 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 20852. 
Publicly available records will be accessible electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

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

    Date of amendment request: December 28, 2000.
    Description of amendment request: The proposed amendment would 
decrease the allowed outage time for an inoperable channel of the 
anticipated transient without scram recirculation pump trip 
instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 9379]]

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 the Technical Specifications are to the 
allowed outage time(s) specified for instrumentation associated with 
the Anticipated Transient Without Scram (ATWS) Reactor Recirculation 
Pump Trip (RPT) system.
    The proposed changes do not involve a change to the plant design 
or operating modes. The changes apply to the ATWS-RPT system, but 
they have no impact on the failure modes or initiators that 
potentially cause an ATWS, and thus have no impact on the frequency 
of occurrence of an ATWS event.
    The proposed changes do not involve a change to the design of 
the ATWS-RPT system, as the proposed changes primarily only affect 
the allowed outage time of the system and do not otherwise affect 
the manner in which the system is tested or operated. Thus, the 
manner in which the ATWS-RPT system is designed to respond to an 
ATWS event is not affected, so its mitigation design function is not 
impacted. Although, by design, on-lime testing of the ATWS-RPT 
requires the system to be rendered unavailable for short periods of 
time, system unavailability is not significantly impacted by the 
proposed changes. The proposed changes involve the establishment of 
a reasonable allowed outage time to support online testing needed to 
periodically confirm system operability, but which minimizes the 
overall system average unavailability. All of the proposed allowed 
outage times are based on the Standard Technical Specifications and 
as such have been determined to be acceptable for maintaining 
adequate ATWS-RPT availability and for minimizing plant risk. They 
thus provide reasonable assurance that the ATWS-RPT system will be 
available on demand to perform its mitigating function in the event 
of an accident or transient involving a failure of the primary scram 
function (i.e., the reactor protection system).
    Based on the above, the proposed changes to the TS do not 
involve a significant increase in the probability or consequences of 
an accident.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes only affect the outage time allowed for the 
ATWS-RPT instrumentation. They do not involve any changes to the 
plant design or operation, and thus do not introduce a new failure 
mode. Therefore, the proposed changes to the TS 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 do not involve a change to the plant design 
or operation including the ATWS-RPT system itself. No change to the 
setpoints of the ATWS-RPT instrumentation is involved. Since ATWS-
RPT availability will be maintained to a sufficiently high degree, 
and since the ATWS-RPT design (including its associated instrument 
setpoints) is unaffected, the TS will continue to provide adequate 
assurance that the ATWS-RPT is capable of performing its intended 
function.
    Based on the above, the proposed changes to the TS do not 
involve a 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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW, Washington, DC 20036-5869.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: December 20, 2000.
    Description of amendment request: The proposed amendment request 
revises Technical Specification (TS) 5.3.1, ``Reactor Core,'' to permit 
the use of the Framatome Cogema Fuels (FCF) ``M5'' advanced alloy for 
fuel rod cladding and fuel assembly spacer grids. The licensee has 
submitted a related exemption request from the requirements of Title 10 
of the Code of Federal Regulations (10 CFR) Section 50.44, ``Standards 
for Combustible Gas Control System in Light-Water-Cooled Power 
Reactors,'' Section 50.46, ``Acceptance Criteria for Emergency Core 
Cooling Systems for Light-Water Nuclear Power Reactors,'' and 
associated Appendix K, ``ECCS Evaluating Models,'' which presume the 
use of zircaloy or ZIRLO cladding. A related Bases change is also made 
to the Bases for TS 2.1.
    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:

    AmerGen has determined that this license amendment request poses 
no significant hazards considerations as defined by 10 CFR 50.92.
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. It has been demonstrated that the material 
properties of the M5 alloy are not significantly different from 
those of Zircaloy-4. Further, there are no evaluated accidents in 
which the fuel cladding or fuel assembly structural components are 
assumed to arbitrarily fail as an accident initiator. The fuel 
handling accident assumes that the cladding does, in fact, fail as a 
result of an undefined fuel handling event. However, the probability 
of that undefined initiating event is independent of the properties 
of the fuel rod cladding. Additionally, in both LOCA [loss-of-
cooling accident] and non-LOCA accident scenarios, there will be no 
significant increase in cladding failure or fission product release, 
since it has been demonstrated that the material properties of the 
M5 alloy are not significantly different from those of Zircaloy-4. 
Therefore, this activity does not involve a significant increase in 
the probability of occurrence or the 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 previously evaluated. It has been 
demonstrated that the material propoerties of the M5 alloy are not 
significantly different than those of Zircaloy-4. Therefore, M5 fuel 
cladding and the fuel assembly structural components will perform 
similarly to those fabricated from Zircaloy-4, thus precluding the 
possibility of the fuel becoming an accident initiator. Therefore, 
this activity does not create the possibility of a new or different 
kind of accident from any 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 material properties of the M5 alloy are not 
significantly different from those of Zircaloy-4 for all normal 
operating and accident scenarios, including both LOCA and non-LOCA 
scenarios * * * Therefore, this activity 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: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: September 14, 2000, as supplemented on 
December 21, 2000.
    Description of amendment request: The licensee proposes to revise 
the

[[Page 9380]]

Technical Specifications to allow a lead fuel assembly (LFA) with a 
limited number of fuel rods clad with advanced zirconium-based alloys 
to be inserted into the core during the next refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies, 
states that fuel rods are clad with either zircaloy or ZIRLO. This 
reflects the requirements of 10 CFR 50.44, 50.46, and 10 CFR Part 
50, Appendix K, which also restricts fuel rod cladding materials to 
zircaloy or ZIRLO. Calvert Cliffs Nuclear Power Plant, Inc. proposes 
to insert a fuel assembly into Calvert Cliffs Unit 2 that have some 
fuel rods clad in zirconium alloys that do not meet the definition 
of zircaloy or ZIRLO. An exemption to the regulations has also been 
requested to allow this fuel assembly to be inserted into Unit 2. 
The proposed change to the Calvert Cliffs Technical Specifications 
will allow the use of cladding materials that are not zircaloy or 
ZIRLO for one fuel cycle once the exemption is approved. To obtain 
approval of new cladding materials, 10 CFR 50.12 requires that the 
applicant show that the proposed exemption is authorized by law, is 
consistent with the common defense and security, will not present an 
undue risk to the public health and safety, and is accompanied by 
special circumstances. The proposed change to the Technical 
Specification is effective only as long as the exemption is 
effective. The addition of what will be an approved temporary 
exemption to Unit 2 Technical Specification 4.2.1 does not change 
the probability or consequences of an accident previously evaluated.
    Supporting analyses indicate that since the LFA will be placed 
in a non-limiting location, the placement scheme and the similarity 
of the advanced alloys to zircaloy-4 will assure that the behavior 
of the fuel rods with these alloys are bounded by the fuel 
performance and safety analyses performed for the zircaloy-4 clad 
fuel rods currently in the Unit 2 core. Therefore, the addition of 
these advanced claddings does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with existing equipment, change the equipment's function, 
or change the method of operating the equipment. The proposed change 
does not affect normal plant operations or configuration. Since the 
proposed change does not change the design, configuration, or 
operation, it could not become an accident initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind] of accident from any previously 
evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The proposed change will add an approved temporary exemption to 
the Unit 2 Technical Specifications allowing the installation of a 
lead fuel assembly. This assembly uses advanced cladding materials 
that are not specifically permitted by existing regulations or 
Calvert Cliffs' Technical Specifications. A temporary exemption to 
allow the installation of this assembly has been requested. The 
addition of an approved temporary exemption to Technical 
Specification 4.2.1 is simply intended to allow the installation of 
the lead fuel assembly under the provisions of the temporary 
exemption. The license amendment is effective only as long as the 
exemption is effective. This amendment does not change the margin of 
safety since it only adds a reference to an approved, temporary 
exemption to the Technical Specifications.
    Supporting analyses indicate that since the LFA will be placed 
in a non-limiting location, the placement scheme and the similarity 
of the advanced alloys to zircaloy-4 will assure that the behavior 
of the fuel rods with these alloys are bounded by the fuel 
performance and safety analyses performed for the zircaloy-4 clad 
fuel rods currently in the Unit 2 core. Therefore, the addition of 
these advanced claddings does not involve a significant reduction in 
the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in [a] margin of safety.

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

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: December 21, 2000.
    Description of amendments request: The amendments would revise 
Technical Specification 5.2.2.e by removing the reference to the 
Nuclear Regulatory Commission (NRC) Policy Statement on working hours.
    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 licensing basis change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change to Technical Specification 5.2.2.e only 
alters the administrative location of, and the regulatory controls 
applicable to, unit staff-specific overtime limits and working 
hours. Overtime limits and working hours will remain controlled by 
plant administrative procedures. Changes to the relocated overtime 
limits and working hours will be controlled in accordance with our 
established procedural control processes. There is no increase in 
the probability of an accident previously evaluated because no 
change is being made to any accident initiator. No previously 
analyzed accident scenario is changed, and initiating conditions and 
assumptions remain as previously analyzed.
    There is no increase in the radiological consequences of any 
accident previously evaluated because the proposed amendment does 
not affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed change does 
not alter the source term, containment isolation, or allowable 
radiological releases.
    Therefore, the proposed amendment does not result in any 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed licensing basis change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed amendment to Technical Specification 5.2.2.e only 
alters the administrative location of and the regulatory controls 
applicable to unit staff specific overtime limits and working hours. 
The proposed amendment does not change the way the plant is 
operated, and no new or different failure modes have been defined 
for any plant system or component important to safety. No limiting 
single failure has been identified as a result of the proposed 
amendment. No new or different types of failures, accident 
initiators or scenarios are introduced by the proposed amendment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed licensing basis change does not involve a 
significant reduction in [a] margin of safety.
    Unit staff overtime is not an input into the calculation of any 
safety margin in the Technical Specification Safety Limits, Limiting 
Safety Settings, or other Limiting Conditions for Operation. Unit 
staff overtime is not an input into the calculation of any safety 
margin in the Technical Requirements Manual, or any other previously 
defined

[[Page 9381]]

margins for any structure, system, or component important to safety. 
The proposed amendment to Technical Specification 5.2.2.e only 
alters the administrative location of, and the regulatory controls 
applicable to unit staff-specific overtime limits and working hours.
    Therefore, the proposed amendment 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 
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: Marsha Gamberoni.

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 application for amendments: December 1, 2000.
    Description of amendments request: The proposed amendments would 
change the Technical Specification (TS) 5.6.3, ``Radioactive Effluent 
Release Report'' date for submittal of the Radioactive Effluent Release 
Report to ``prior to May 1'' of each year.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change is administrative in nature. The date of 
submittal of the Radioactive Effluent Release Report is not an 
initiator of any analyzed event. Similarly, the date of submission 
does not affect the consequences of any accident previously 
evaluated. The proposed change will not physically alter the plant, 
and it will not affect plant operation. The proposed change to the 
submission date of the Radioactive Effluent Release Report will 
continue to meet the reporting requirement of 10 CFR 50.36a(a)(2) 
and further clarifies when the report is to be submitted. As such, 
the proposed change does not involve an increase in the probability 
or consequence of any accident previously evaluated.
    2. The proposed license amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed TS change is administrative in nature. It would 
revise the date by which the Radioactive Effluent Release Report is 
required to be submitted to the NRC. Revision of the submittal date 
for the report will not affect any accident initiator or cause any 
new accident precursors to be created. The proposed change will not 
affect the types or amounts of radioactive effluents released or 
cumulative occupational radiological exposures.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to the submittal requirement for the 
Radioactive Effluent Release Report is only an administrative change 
and will have no [effect] on any 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 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards considerations.
    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: December 29, 2000.
    Description of amendment request: The proposed amendment would 
modify the applicability statements of Technical Specification (TS) 
Limiting Conditions for Operations (LCOs) 3.3.6.2, ``Secondary 
Containment Isolation Instrumentation,'' 3.3.7.1, ``Control Room 
Emergency Filtration (CREF) System Instrumentation,'' 3.6.4.1, 
``Secondary Containment,'' 3.6.4.2, ``Secondary Containment Isolation 
Valves (SCIVs),'' 3.6.4.3, ``Standby Gas Treatment (SGT) System,'' 
3.7.3, ``Control Room Emergency Filtration (CREF) System,'' 3.7.4, 
``Control Center Air Conditioning (AC) System,'' 3.8.2, ``AC Sources--
Shutdown,'' 3.8.5, ``DC Sources--Shutdown,'' and 3.8.8, ``Distribution 
Systems--Shutdown.'' The proposed modifications would require 
operability of the associated systems only if recently irradiated fuel, 
which is identified as fuel that has occupied part of a critical 
reactor core within the previous 7 days, is handled during the first 
few days of an outage. The 7-day value is based on the results of a 
revised analysis of a fuel handling accident (FHA) that was performed 
by utilizing the guidelines contained in NRC Regulatory Guide 1.183, 
``Alternative Radiological Source Terms for Evaluating Design Basis 
Accidents at Nuclear Power Reactors,'' dated July 2000.
    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 change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The new ``recently irradiated fuel'' term to describe irradiated 
fuel assemblies is used to establish operational conditions where 
specific activities represent situations where significant 
radioactive releases can be postulated. These operational conditions 
are consistent with the design basis analysis. Because the equipment 
affected by the revised operational conditions is not an initiator 
to any previously analyzed accident, the proposed change cannot 
increase the probability of any previously evaluated accident.
    The re-analysis of the Fuel Handling Accident concludes that 
radiological consequences are within the acceptance criteria in 
Regulatory Guide 1.183 (Reference 3 [of the licensee's application 
dated December 27, 2000]). The results of the Core Alterations 
events other than the Fuel Handling Accident remain unchanged from 
the original design basis, which showed that these events do not 
result in fuel cladding damage or radioactive release. The FHA re-
analysis includes a drop of a non-irradiated fuel assembly over 
recently irradiated assemblies in the reactor core 24 hours after 
reactor shutdown. The radiological consequences associated with this 
scenario, assuming no mitigation credit for Secondary Containment, 
SGT and CREF Systems, have been shown to satisfy the acceptance 
criteria in Reference 3. Therefore, the proposed changes do not 
significantly increase the radiological consequences of any 
previously evaluated accident.
    Based on the above, the proposed changes do not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed requirements are imposed when specific activities 
represent situations where significant radioactive releases are not 
postulated. The proposed requirements are supported by the revised 
design basis Fuel Handling Accident analysis. The proposed changes 
do not introduce any new modes of plant operation and do not involve 
physical modifications to the plant. Therefore, the proposed change 
does not create the potential for 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.

[[Page 9382]]

    The proposed changes revise the Fermi 2 TS[s] to establish 
operational conditions where specific activities represent 
situations during which significant radioactive releases can be 
postulated. These operational conditions are consistent with the 
design basis analysis and are established such that the radiological 
consequences are at or below the regulatory guidelines. Safety 
margins and analytical conservatisms are retained to ensure that the 
analysis adequately bounds all postulated event scenarios. The 
proposed TS Applicability statements continue to ensure that the 
TEDE [total effective dose equivalent] at both the Control Room and 
the exclusion area and low population zone boundaries are below the 
corresponding regulatory guidelines in Reference 3 [of the 
licensee's application dated December 27, 2000]; therefore, the 
proposed change will not result in 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: Claudia M. Craig.

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: December 28, 2000
    Description of amendment request: The proposed amendments would 
revise the Technical Specification requirements associated with storage 
of spent fuel in the spent fuel storage pools to account for 
degradation of the Boraflex panels used in the construction of the 
storage racks and maintain acceptable margins of subcriticality in the 
spent fuel storage pools.
    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 change involve a significant increase in the 
probability or consequence of an accident previously evaluated?
    Response: No.
    The proposed Oconee Nuclear Station (ONS) Technical 
Specification (TS) changes described in the License Amendment 
Request (LAR) do not create a significant increase in the 
probability or consequence of an accident previously evaluated.
    The loss of boron from the Boraflex panels in the Spent Fuel 
Pool (SFP) racks is offset by the presence of soluble boron in the 
SFP water for criticality control. The increased surveillance 
frequency provides assurance the SFP boron concentration limits will 
be maintained. The handling of the fuel assemblies in the SFP has 
always been performed in borated water. Fuel assembly placement in 
the revised fuel storage configurations described in the LAR will 
continue to be controlled by approved fuel handling procedures to 
ensure compliance with TS requirements.
    The proposed changes do not affect the probability of a dropped 
fuel assembly accident, accidental misloading of spent fuel, or 
heavy load drop onto the SFP racks. The criticality analyses show 
the consequences of such events are not affected by the proposed 
changes and that the fuel will remain subcritical.
    The radiological consequences of a fuel misloading or handling 
accident in the SFP, or a heavy load drop onto the SFP racks, do not 
change by taking credit for soluble boron in the pool because the 
current SFP boron concentration limit is unchanged.
    In the unlikely event of significant SFP temperature increases 
or decreases, the proposed soluble boron limits and increased 
surveillance frequency of the SFP boron concentration provide 
assurance the fuel will remain subcritical.
    2. Will the change create the possibility of a new or different 
kind of accident from any previously evaluated?
    Response: No.
    Criticality accidents in the SFP are not new or different types 
of accidents. They have been analyzed as described in Section 
9.1.2.3.2 of the Updated Final Safety Analysis Report and in 
Criticality Analysis reports associated with specific licensing 
amendments for fuel enrichments up to 5.00 weight percent U-235. The 
evaluations described in the LAR demonstrate that the proposed 
changes do not create a new or different kind of accident from any 
previously analyzed. The accident analysis in the Updated Final 
Safety Analysis Report remains bounding.
    There are no changes in equipment design or in plant 
configuration. The revised requirement will not result in the 
installation of any new equipment or modification of any existing 
equipment. Therefore, the proposed changes will not result in the 
possibility of a new or different kind of accident.
    3. Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed TS changes and the resulting spent fuel storage 
operating limits provide adequate safety margin to ensure that the 
stored fuel assembly array will always remain subcritical. Those 
limits are based on the ONS spent fuel pool-specific criticality 
analyses described in the LAR.
    The criticality analyses are based on the methodology described 
in WCAP-14416-NP-A, ``Westinghouse Spent Fuel Rack Criticality 
Analysis Methodology,'' Revision 1, November 1996, which has been 
reviewed and approved by the NRC. This methodology takes partial 
credit for soluble boron in the SFP and meets the following NRC 
acceptance criteria (10 CFR 50.68) for preventing criticality 
outside the reactor:
    a. keff shall be less than 1.0 if fully flooded with 
unborated water, which includes an allowance for uncertainties at a 
95% probability, 95% confidence (95/95) level; and
    b. keff shall be less than or equal to 0.95 if fully 
flooded with borated water, which includes an allowance for 
uncertainties at a 95/95 level.
    The proposed TS limits provide a level of safety comparable to 
the conservative criticality analysis methodology required by USNRC 
Standard Review Plan for the Review of Safety Analysis Reports for 
Nuclear Power Plants, LWR Edition, NUREG-0800, June 1987, USNRC 
Spent Fuel Storage Facility Design Bases (for comment) Proposed 
Revision 2, 1981, Regulatory Guide 1.13, and ANSI/ANS-57.2-1983.
    Therefore, the proposed changes will not result in a significant 
reduction in the plant's margin of safety.
    Based on the above evaluations, Duke concludes that the 
activities associated with the above described changes present no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92 and accordingly, a finding by the NRC of no significant 
hazards consideration is justified.

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

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

    Date of amendment request: July 21, 2000, as supplemented by letter 
dated December 13, 2000.
    Description of amendment request: The proposed amendment would 
reduce the limit for reactor coolant system (RCS) specific activity in 
technical specification (TS) 3/4.4.8. The dose equivalent iodine 131 
(I-131) is proposed to be lowered from the current value of  
0.35 micro Curies per gram (Ci/gram) to a value of  
0.20 Ci/gram as specified in TS 3.4.8.a (and associated 
Actions and Table 4.4-12). This change will also lower the 
``''Acceptable Operation'' line on Figure 3.4-1 from 21 Ci/
gram to 12 Ci/gram Dose Equivalent I-131 for 80-percent to 
100-percent power, and a commensurate reduction for power between 20-
percent and 80-percent power.
    In conjunction with the reduced TS limit for RCS specific activity, 
the

[[Page 9383]]

Beaver Valley Power Station (BVPS) Unit 1, control room and offsite 
dose consequences resulting from a postulated Main Steam Line Break 
have been re-analyzed to allow for higher primary-to-secondary leakage 
in accordance with methodology described in Nuclear Regulatory 
Commission (NRC) Generic Letter (GL) 95-05, ``Voltage-Based Repair 
Criteria for Westinghouse Steam Generator Tubes by Outside Diameter 
Stress Corrosion Cracking,'' and as previously approved in BVPS-1 
license amendment number 205.
    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?
    The proposed change, which lowers the Technical Specification 
limit for Dose Equivalent I-131, is conservative and will not 
adversely affect the current calculated dose values for BVPS Unit 1 
Design Basis Accidents (DBAs) since a lower RCS specific activity 
will lower the calculated dose from any resultant steam generator 
tube leakage postulated during the DBA. The Standard Review Plan 
assumption for accident-induced steam generator tube leakage spike 
remains valid. Thus, the dose listed in the BVPS Unit 1 UFSAR 
[Updated Final Safety Analysis Report] from those DBAs which 
calculate and list a dose value in the BVPS Unit 1 UFSAR will remain 
bounding values, except for the Main Steam Line Break (MSLB) DBA.
    The immediate effect upon receiving a revised lower primary 
coolant specific activity limit in Technical Specification 3.4.8.a 
would also result in a lower calculated MSLB dose value, if 
incorporated into the MSLB dose calculation without any other 
modifications. But the BVPS Unit 1 MSLB analysis is analyzed per GL 
95-05 which states that a reduction [in] RCS iodine activity is an 
acceptable means for accepting higher projected leakage rates and 
still meeting the applicable limit of Title 10 of the Code of 
Federal Regulations Part 100 and GDC [General Design Criterion] 19 
utilizing currently accepted licensing basis assumptions. Thus, 
pursuant to this GL 95-05 methodology, the reduced RCS specific 
activity limit for Technical Specification 3.4.8.a will be used to 
allow for higher projected leakage rates, while still meeting the 
applicable regulatory dose limits.
    Thus, the current BVPS Unit 1 MSLB calculated dose value will 
not decrease with a new lower RCS specific activity value in order 
to allow for a higher projected leakage rates[sic]. However, the 
BVPS Unit 1 MSLB calculated dose values will remain within the 
limits specified in 10 CFR 50 Appendix A, GDC 19, and the 
radiological doses to the public will remain a small fraction of the 
regulatory limits specified in 10 CFR 100.11, using methodology 
previously accepted in BVPS Unit 1 License Amendment No. 205.
    Therefore, this proposed change will not increase the 
probability of occurrence of a postulated accident or will not 
significantly increase the consequences of an accident previously 
evaluated since the change would continue to comply with the current 
BVPS Unit 1 and Unit 2 licensing basis as it relates to the dose 
limits of GDC 19 and 10 CFR Part 100.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed license amendment to the primary coolant specific 
activity limit does not change the way the RCS is operated. The 
proposed changes only involve changes to the primary coolant 
specific activity limit where continued power operation may occur. 
This reduced limit is conservative and does not alter the RCS or 
steam generators' ability to perform their design bases [functions].
    GL 95-05 states that any reduction of RCS specific activity less 
than 0.35 Ci/gram Dose Equivalent I-131 requires an 
evaluation of release rate data. This evaluation shows that BVPS 
Unit 1 RCS Dose Equivalent I-131 data fully supports lowering the 
Technical Specification RCS specific activity limit to 0.20 
Ci/gram without compromising the Standard Review Plan 
assumption of a post-event iodine spike factor of 500.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any previously evaluated 
accident since the RCS and steam generator will continue to operate 
in accordance with their design bases.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not adversely affect the ability 
of systems, structures or components important to the mitigation and 
control of design bases accident conditions within the facility. In 
addition, the proposed amendment does not affect the ability of 
safety systems to ensure that the facility can be maintained in a 
shutdown or refueling conditions for extended periods of time.
    The proposed license amendment to the primary coolant specific 
activity limit does not adversely change the way the RCS or steam 
generators are operated. This modification does not alter these 
systems' ability to perform their design bases [functions]. The 
existing safety analyses remain bounding. Therefore, the margin of 
safety is not significantly 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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Marsha Gamberoni.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of amendment request: December 4, 2000.
    Description of amendment request: The proposed license amendment 
would revise the St. Lucie Unit 1 Updated Safety Analysis Report to 
reflect the new main steam line break (MSLB) analysis treatment of a 
hypothesized single failure of a main feedwater isolation valve (MFIV). 
The new analysis of the MSLB terminates feedwater addition to the 
faulted steam generator by crediting MFIV closure and tripping the main 
feedwater (MFW) and condensate pumps.
    This proposed change to the Unit 1 licensing bases for the MSLB 
analysis is required to resolve an existing Generic Letter 91-18 
degraded, but operable, condition regarding the postulated peak 
pressure during an MSLB inside containment. In December 1998 the draft 
results of a Unit 1 MSLB containment re-analysis indicated an 
unexpected higher peak containment pressure of 55.9 psig. The Unit 1 
containment design pressure is 44 psig. The cause for the higher peak 
pressure in the re-analyzed MSLB event is that non-conservative 
assumptions were used in the original analysis of record. When these 
non-conservatisms were corrected and input to the MSLB licensing bases 
analysis, the containment peak pressure exceeded the containment design 
pressure. This condition was reported to the NRC via Licensee Event 
Report No. 50-335/1998-009.
    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.
    This amendment changes the licensing bases for the MSLB analysis 
to credit a trip of the non-safety MFW and condensate pumps as a 
backup method to terminate feedwater addition should a MFIV fail to 
close. This activity has no increase in the probability of a MSLB, 
as no physical changes are being made to the steam generators, main 
steam piping, and the normal operating temperatures and pressures 
for the main steam system remain

[[Page 9384]]

unchanged. This activity also has no adverse effect on the 
consequences of an accident because the MSLB containment response is 
bounded by the new analysis. Main feedwater termination occurs 
during a postulated MSLB such that the containment design pressure 
is not exceeded. Although a circuit failure (short) in the MSIS 
[main steam isolation signal] backup trip of the MFW and condensate 
pump breakers would result in tripping the running MFW and 
condensate pumps, this is less probable due to the energized to 
actuate design than existing postulated failures in the MSIS 
circuitry that would also lead to a loss of feedwater event. 
Therefore, 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.
    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.
    This amendment changes the licensing bases for the MSLB analysis 
to credit a trip of the non-safety MFW and condensate pumps as a 
backup method to terminate feedwater addition should a MFIV fail to 
close. The physical modifications made to support the installation 
of the new pneumatic valve operators for the MFIVs and installation 
of the backup main steam isolation signal (MSIS) trip of the non-
safety MFW and condensate pumps conform to all applicable design 
standards. Failure modes introduced by these changes are bounded by 
the original design, and no other physical changes were made to the 
plant. Therefore, 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.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    This amendment changes the licensing bases for the MSLB inside 
containment response analysis to credit a trip of the non-safety MFW 
and condensate pumps as a backup method to terminate feedwater 
addition should a MFIV fail to close. This differs from the 
currently licensed analysis that credits the closure of redundant 
safety related valves for main feedwater termination single failure 
considerations. However, Sections 6.2.1.4 and 15.1.5 of the Standard 
Review Plan allows the use of a non-safety backup in response to a 
failure of safety related components with regards to mitigating the 
effects of the mass energy release of ruptured secondary piping 
inside containment. This change to the licensing bases is consistent 
with the guidance provided in the Standard Review Plan. In addition, 
a probabilistic safety assessment was performed to evaluate the 
change in main feedwater isolation reliability between crediting 
redundant safety related isolation valves or safety related 
isolation valves and trip of the non-safety MFW and condensate 
pumps. This assessment concluded that the change in reliability is 
not risk significant. Therefore, operation of the facility in 
accordance with the proposed amendment would 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 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 No. 50-389, St. Lucie Plant, 
Unit No. 2, St. Lucie County, Florida

    Date of amendment request: November 28, 2000.
    Description of amendment request: The proposed license amendment 
would revise the Updated Final Safety Analysis Report (UFSAR) with a 
revised post-trip steam line break (SLB) analysis. The design basis for 
the current analysis of record ensures that no fuel failure will occur 
for all post-trip SLB cases. The new analysis supports a change to the 
fuel failure criterion, to limit fuel failure to less than or equal to 
2%. The change in allowed fuel failure fraction results in a shutdown 
margin benefit and provides additional flexibility in the core design. 
Limits for the physics parameters that most affect the post-trip SLB 
results will be established on a core-specific basis and included in 
the Core Operating Limits Report for each cycle. The revised analysis, 
with the limit of 2% fuel failure, continues to meet the 10 CFR part 
100 dose criteria.
    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 proposed amendment revises the post-trip SLB analysis of 
record to support a fuel failure limit of 2% as compared to the 
current criterion of no fuel failure. Post-trip SLB is a current 
design basis event for St. Lucie Unit 2 and is defined in the St. 
Lucie Unit 2 Updated Final Safety Analysis Report (UFSAR). The 
revision to the analysis does not impact the event initiator and 
requires no change to any plant component or system. The plant 
configuration remains unchanged, and thus the probability of 
occurrence of previously analyzed accidents is not affected by the 
proposed change.
    Radiological consequences for the return to power (RTP) SLB 
event for St. Lucie Unit 2 have been calculated to infer the allowed 
fuel failure fraction from the 2-hour and 8-hour 10 CFR 100 dose 
limits and are consistent with the results presented in License 
Amendment 105. Releases were calculated based on fuel that violates 
Centerline-Melt (CTM) criteria and produces fuel failure limits of 
13.5% for inside containment SLB and 3.4% fuel failure for an 
outside containment SLB.
    These fuel failure values represent an upper bound limit 
corresponding to the 10 CFR 100 dose criteria. A conservative value 
of 2% fuel failures (from violation of CTM and/or departure nucleate 
boiling ratio (DNBR) specified acceptable fuel design limits 
(SAFDL)) will be utilized as a cycle specific limit for post-trip 
SLB. The peak power density during the post-trip SLB also will be 
limited to less than or equal to 30 kW/ft. For each fuel cycle core 
design, these limits will be verified based on the calculated 
physics data for that cycle.
    The limit of 2% fuel failures, in conjunction with the 30 kW/ft 
on peak power density, ensures a coolable geometry during and 
subsequent to the post-trip SLB RTP. This fuel failure limit, along 
with a conservative allowance for DNB propagation failures, remains 
well below the upper bound limits of 13.5% and 3.4% fuel failure for 
the inside and the outside containment breaks, respectively, 
corresponding to the 10 CFR 100 dose criteria.
    Therefore, 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.
    2. Use of the proposed amendment would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed amendment is merely a revision to the post-trip SLB 
event analysis, which continues to meet the applicable limits of 10 
CFR 100 dose criteria. There is no change to the plant 
configuration, systems, or components that would create new failure 
modes. The modes of operation of the plant remain unchanged.
    Therefore, 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.
    3. Use of the proposed amendment would not involve a significant 
reduction in a margin of safety.
    The proposed amendment revises the post-trip SLB analysis and 
supports a change to the fuel failure acceptance criterion. The 
revised analysis, with the limit of 2% fuel failure, would continue 
to provide margin to the applicable limits of 10 CFR 100 dose 
criteria. The proposed change, including any core design variations, 
will have no adverse impact on other plant safety analysis. The 
plant operation would continue to remain within all design basis 
requirements, which would ensure that a safety margin to the

[[Page 9385]]

acceptance criteria would continue to remain available during plant 
operation at all power levels.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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 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: December 6, 2000.
    Description of amendment request: The proposed license amendments 
would revise the Turkey Point Units 3 and 4 Technical Specification 
(TS) Surveillance Requirement (SR) 4.6.1.3.c to allow performance of 
the required surveillance test of the air lock interlock at an interval 
of 24 months. Currently, SR 4.6.1.3.c requires the interlocks to be 
tested at least once every six month, and is, therefore, done with the 
plant online.
    Each containment at Turkey Point has two air locks, commonly named 
the personnel air lock and the escape hatch. Each air lock has an inner 
and an outer door. Interlocks prevent both doors in the air lock from 
being opened at the same time, thereby preserving containment 
integrity, when required. These interlocks are completely mechanical, 
and contain no degradable components. Historically, the air lock 
interlock test frequency was chosen to coincide with that of the 
overall airlock leakage test. Turkey Point Units 3 and 4 TSs were 
amended in January 1997, to permit the extension of the overall airlock 
leakage test frequency up to a maximum of 30 months, based on 
acceptable test results. The licensee requested revision of SR 
4.6.1.3.c. to require testing of the air lock interlock at an interval 
of 24 months, which would also allow a maximum interval of up to 30 
months between tests. Therefore, the proposed amendments would realign 
the SR frequencies of the air lock interlock test and the overall 
leakage test with each other. In support of these amendments, the 
licensee stated that currently the SR test is being performed with the 
plant online, when the interlocks are required to be operable. If the 
proposed amendments are granted, the licensee expects to perform the 
test during refueling outages, when the plant is in a mode in which the 
interlock is not required to be operable. Also, the licensee stated 
that the proposed amendments are consistent with the as-low-as-
reasonably-achievable principles, because they would preclude 
performance of the test with the plant online, which involves some risk 
of dose to workers.
    Additionally, the licensee requested to amend TS 3.3.2, Table 3.3-
2, Item 1.e, that addresses the requirements for the safety injection 
signal (SIS) generated by high steamline differential pressure, to 
change the asterisk following Modes 1, 2, 3, to a pound sign (i.e., #). 
The licensee stated that the existing asterisk refers to an incorrect 
note, in that it indicates that the SIS may be blocked below the Tavg--
Low Interlock Setpoint, when in fact the Block Permissive for this SIS 
is pressurizer pressure below 2000 psi. The licensee stated that this 
is due to a typographical error, and that the change is requested to 
make the Mode Applicability consistent with the design of the 
protection logic.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes allow performance of the required 
surveillance at the same frequency as the performance of the air 
lock overall leakage surveillance. The proposed relaxation in 
surveillance frequency will not impact the initiating event for any 
previously evaluated accident. The correction of the typographical 
error has no impact on any accident analysis. The proposed changes 
do not affect any of the assumptions made or methodologies used for 
any accident analysis. Thus the proposed changes have no impact on 
any of the accident probabilities or consequences. Therefore, the 
proposed amendments do 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 changes do not alter the design, physical 
configuration, or modes of operation of the plant. No changes are 
being made to the plant that would introduce any new accident causal 
mechanisms. The proposed Technical Specification changes do not 
impact any other plant systems. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes do not change the operation, function, or 
modes of plant or equipment operation. The proposed changes do not 
change the level of assurance of containment integrity. Plant 
processes and training preclude challenges to the air lock 
interlocks. The correction of the typographical error has no impact 
on any margin of safety. Therefore, operation of the facility in 
accordance with the proposed amendments would 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 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.

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

    Date of amendment request: December 5, 2000.
    Description of amendment request: The proposed amendment would 
implement programmatic controls for radiological effluent technical 
specifications (RETS) in the administrative section of the Technical 
Specifications (TSs) and relocate the procedural details of the RETS to 
the offsite dose calculation manual (ODCM), the process control program 
(PCP), or other new programs, consistent with the guidance of Standard 
TSs (STS) (NUREG-1433) and NRC Generic Letter 89-01.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 9386]]

    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes are administrative in nature and alter only the 
format and location of programmatic controls and procedural details 
relative to radioactive effluents, radiological environmental 
monitoring, radioactive source leakage testing, solid radioactive 
wastes, and associated reporting requirements. Existing TS 
containing procedural details on radioactive effluents, radiological 
environmental monitoring, radioactive source leakage testing, 
explosive gas monitoring, storage tank radioactive content limits, 
solid radioactive wastes and associated reporting requirements are 
being relocated to the ODCM, PCP or other new programs as 
appropriate. Compliance with applicable regulatory requirements will 
continue to be maintained. In addition, the proposed changes do not 
alter the conditions or assumptions in any of the previous accident 
analyses. Since the previous accident analyses remain bounding, the 
radiological consequences previously evaluated are not adversely 
affected by the proposed changes.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes do not involve any change to the configuration 
or method of operation of any plant equipment. Accordingly, no new 
failure modes have been defined for any plant system or component 
important to safety nor has any new limiting single failure been 
identified as a result of the proposed changes. Also, there will be 
no change in types or increase in the amounts of any effluents 
released offsite.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes do not involve any actual 
change in the methodology used in the control of radioactive 
effluents, radioactive sources, solid radioactive wastes, or 
radiological environmental monitoring. These changes are considered 
administrative in nature and provide for the relocation of 
procedural details outside of the technical specifications but add 
appropriate administrative controls to provide continued assurance 
of compliance to applicable regulatory requirements. These proposed 
changes also comply with the guidance contained in Generic Letter 
89-01 and the STS.
    Therefore, it can be concluded a significant reduction in the 
margin of safety would not be involved.

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

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

    Date of amendment request: January 10, 2001.
    Description of amendment request: The proposed amendment would 
remove the standby liquid control pump flow surveillance requirement to 
recycle demineralized water to the test tank and change the testing 
frequency from monthly to quarterly.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The only significant consequence of these changes compared to 
present plant operation will be to change the test frequency of the 
MNGP [Monticello Nuclear Generating Plant] SLC [standby liquid 
control] pump capacity test to quarterly, which has been previously 
reviewed and approved by the NRC staff for similar boiling water 
reactors (BWRs). There are no changes to equipment performance or 
postulated failure modes. The change does not affect the assumptions 
or methods of accident mitigation previously evaluated. The proposed 
amendment will have no impact on the probability or consequences of 
an accident.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The only significant consequence of these changes compared to 
present plant operation will be to change the test frequency of the 
MNGP SLC capacity flow test to quarterly, which has been previously 
reviewed and approved by the NRC staff for similar BWRs. The change 
does not affect or introduce any new plant operating modes. The 
changes do not alter any existing system interaction and do not 
introduce any new failure modes. The proposed amendment will not 
create the possibility for any new or different accidents for those 
previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The only significant consequence of these changes compared to 
present plant operation will be to change the test frequency of the 
MNGP SLC pump capacity test to quarterly, which has been previously 
reviewed and approved by the NRC staff for similar BWRs. There is no 
change in the reliability or performance of the SLC system. Other 
surveillance requirements assure that SLC hydraulic conditions will 
not degrade between quarterly surveillances. The proposed changes 
have no effect on the mitigation of any postulated accident or event 
at MNGP. The proposed Technical Specification 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 E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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 20, 2000.
    Description of amendment request: The proposed amendments will 
implement changes to the Technical Specifications to increase the 
allowable deviation in individual rod position indication (IRPI). The 
portion of this amendment that pertains to control rod misalignment 
above 85 percent as a function of peaking factors will be reviewed by 
the NRC staff as a separate action.
    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.
    Based on the analyses documented in WCAP-15432, Revision 1, all 
pertinent licensing-basis acceptance criteria have been met and the 
margin of safety, as defined in the Technical Specification Bases, 
is not significantly reduced in any of the Point Beach licensing 
basis accident analyses based on the subject change. Therefore, the 
probability of an accident previously evaluated has not 
significantly increased. Because design limitations continue to be 
met and the integrity of the reactor coolant

[[Page 9387]]

system pressure boundary is not challenged, the assumptions employed 
in the calculation of the offsite radiological doses remain valid. 
Neither rod position indication nor the limits on allowed rod 
position deviation is an accident initiator or precursor. Therefore, 
the consequences of an accident previously evaluated will not be 
significantly increased.
    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.
    Based on the analyses documented in WCAP-15432, Revision 1, all 
pertinent licensing-basis acceptance criteria have been met and the 
margin of safety, as defined in the Technical Specification Bases, 
is not significantly reduced in any of the Point Beach licensing 
basis accident analyses based on the subject change.
    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 changes described in the amendment are supported by 
the analyses and evaluations described in Attachment 2 of this 
letter (safety evaluation) [licensee's application dated November 
20, 2000]. 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.
    All equipment important to safety will continue to operate as 
designed. 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.
    Based on the analyses documented in WCAP-15432, Revision 1, all 
pertinent licensing-basis acceptance criteria have been met and the 
margin of safety, as defined in the Technical Specification Bases, 
is not significantly reduced in any of the Point Beach licensing 
basis accident analyses based on the subject changes to safety 
analyses input parameter values. There are no new or significant 
changes to the initial conditions contributing to accident severity 
or consequences. Since the safety evaluation in Attachment 2 of this 
letter [licensee's application dated November 20, 2000] demonstrates 
that all applicable acceptance criteria continue to be met, the 
subject operating conditions will not involve a significant 
reduction in a margin of safety at Point Beach.

    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: Claudia M. Craig.

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.
    Description of amendment request: The proposed changes would revise 
the Technical Specifications and Technical Requirements Manual 
requirements applicable when actions direct suspension of operations 
involving positive reactivity changes, by removing the requirement not 
to make positive reactivity changes during certain plant conditions, 
and by limiting the amount of reactivity changes that are allowed to 
those that will continue to assure appropriate reactivity limits are 
met. Related changes to the Bases are also proposed. In addition, an 
administrative change is also proposed to remove a footnote that 
allowed an alternate onsite emergency power source to be substituted 
for one of the required diesel generators for 21 consecutive days for 
refueling outages 1RE05 and 2RE04 only.
    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?
    The proposed change does not involve an increase in the 
probability or consequences of an accident previously evaluated. The 
proposed activities to be allowed during certain operating 
conditions are permitted at other times during routine operating 
conditions. The changes do not affect the limits on reactivity that 
are specified in other specifications. The proposed changes do not 
reduce restrictions on addition or flowpaths of unborated water that 
are in the existing specifications. The proposed change does not 
affect the limits on reactivity that are credited in the safety 
analysis. Therefore, no increase in the probability or consequences 
of any accident previously evaluated will occur.
    In addition to the changes proposed to controls over reactivity 
changes, an administrative change is proposed to remove a footnote 
that is no longer applicable to the facility. Since the footnote no 
longer has meaning or relevance to the operation of the facility, 
its removal does not increase the probability or consequences of any 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes merely permit the conduct of normal 
operating evolutions during limited periods when additional controls 
over reactivity margin are imposed by the Technical Specifications. 
The proposed change does not introduce any new equipment into the 
plant or significantly alter the manner in which existing equipment 
will be operated. The changes to operating allowances are minor and 
are only applicable during certain conditions. The operating 
allowances are consistent with those acceptable at other times. 
Since the proposed changes only allow activities that are presently 
approved and routinely conducted, no possibility exists for a new or 
different kind of accident from those previously evaluated.
    In addition to the changes proposed to controls over reactivity 
changes, an administrative change is proposed to remove a footnote 
that is no longer applicable to the facility. Since the footnote no 
longer has meaning or relevance to the operation of the facility, 
its removal cannot create the possibility of a new or different kind 
of accident from those previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes do not involve a significant reduction in a 
margin of safety because the ability to make the reactor subcritical 
and maintain it subcritical during all operating conditions and 
modes of operation will be maintained. The margin of safety is 
defined by the shutdown margin limits and the refueling boron 
concentration limit. The proposed changes do not affect these 
operating restrictions and the margin of safety which assures the 
ability to make and maintain the reactor subcritical is not 
affected.
    In addition to the changes proposed to controls over reactivity 
changes, an administrative change is proposed to remove a footnote 
that is no longer applicable to the facility. Since the footnote no 
longer has meaning or relevance to the operation of the facility, 
its removal cannot result in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 6, 2000.

[[Page 9388]]

    Description of amendment request: The proposed amendment would 
revise each of the three units' Technical Specifications (TS) to 
provide action requirements and completion times for use under plant 
conditions involving one inoperable low pressure coolant injection 
(LPCI) pump in each of the two emergency core cooling system (ECCS) 
divisions. The new requirements are consistent with those currently 
specified for use under conditions of two inoperable LPCI pumps in the 
same ECCS division.
    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's analysis is presented below:
    A. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The LPCI system consists of two independent LPCI subsystems. Each 
of the two LPCI subsystems has two LPCI pumps. The current TS Limiting 
Conditions for Operation are based on ECCS analyses that postulate the 
failure of one entire subsystem. New analyses have been performed that 
postulate the failure of one LPCI pump in each subsystem (i.e., both 
subsystems operating at reduced capacity). The new analyses show that 
the total ECCS flow capacity provided by the entire LPCI system is 
greater when operating under the newly-analyzed conditions than for the 
previously analyzed conditions. Thus, the action requirements and 
completion times associated with inoperability of two LPCI pumps in on 
the same LPCI subsystem may be applied in cases when one LPCI pump is 
inoperable in each LPCI subsystem. Since ECCS performance is not 
adversely affected, there is no increase in the probability or 
consequences of any analyzed accident.
    B. The proposed amendment 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. The proposed 
change will not impose any new or eliminate any existing requirements.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    The proposed change will not reduce a margin of safety because it 
has no adverse effect on any safety analyses assumptions. The proposed 
new Conditions involving one inoperable LPCI pump in each LPCI 
injection subsystem represent more reliable configurations than the 
existing LCOs which apply for two inoperable LPCI pumps in one ECCS 
subsystem.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of application for amendments: January 22, 2001 (TS 00-01).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah Nuclear Plant Technical Specification surveillance 
requirements for assuring against ice condenser flow blockage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The only analyzed accidents of possible consideration in regards 
to changes potentially affecting the ice condenser are a loss-of-
coolant accident (LOCA) and a high energy line break (HELB) inside 
containment. However, the ice condenser is not postulated as being 
the initiator of any LOCA or HELB. This is because it is designed to 
remain functional following a design basis earthquake, and the ice 
condenser does not interconnect or interact with any systems that 
interconnect or interact with the reactor coolant or main steam 
systems.
    Neither the TS [Technical Specification] amendment nor the TS 
Bases changes can increase the probability of occurrence of any 
analyzed accident because they are not the result or cause of any 
physical modification to ice condenser structures, and for the 
current design of the ice condenser, there is no correlation between 
any credible failure of it and the initiation of any previously 
analyzed event.
    Regarding the consequences of analyzed accidents, the ice 
condenser is an engineered safety feature designed, in part, to 
limit the containment subcompartment and steel containment vessel 
pressures immediately following the initiation of a LOCA or HELB. 
Conservative subcompartment pressure analysis shows this criteria 
will be met if the reduction in the flow area per bay provided for 
ice condenser air and or steam flow channels is less than or equal 
to 15 percent, or if the total flow area blocked within each lumped 
analysis section is less than or equal to the 15 percent as assumed 
in the safety analysis.
    The proposed amendment also revises the flow area verification 
surveillance frequency from at least once per 12 months to at least 
once per 18 months such that it will coincide with refueling 
outages. Management of ice condenser maintenance activities has 
successfully limited activities, with the potential for significant 
flow channel degradation, to the refueling outage. Verifying an ice 
bed is left with less than or equal to 15 percent flow channel 
blockage at the conclusion of a refueling outage assures the ice bed 
will remain in an acceptable condition for the duration of the 
operating cycle. During the operating cycle, a certain amount of ice 
sublimates and reforms as frost on the colder surfaces in the ice 
condenser. However, frost does not degrade the flow channel flow 
area. The surveillance will effectively demonstrate operability for 
an allowed 18-month surveillance period. Therefore, increasing the 
surveillance interval does not affect the ice condenser operation or 
accident response. Limiting ice bed flow channel blockage to less 
than or equal to 15 percent ensures operation is consistent with the 
assumptions of the DBA analyses. Thus, the proposed amendment for 
flow blockage determination provides the necessary assurance that 
flow channel requirements are met without additional evaluations and 
thus will not increase the consequences of a LOCA or HELB.
    In regard to [the] TS 3.6.5.3 Bases change, clarifying the 
action entry of Action b to not apply when personnel are standing on 
or opening doors for a short duration to perform surveillances or 
minor maintenance activities, such as ice removal, does not increase 
analyzed accident consequences. These are not new or additional 
actions compared to those performed previously, the probability of 
an accident versus the time to perform these actions is small, the 
number of personnel involved is small, and their duration is 
generally much less than the four-hour frequency of required Action 
b (monitor maximum ice condenser temperature). Therefore, these 
activities do not adversely affect ice bed sublimation, melting, or 
ice condenser flow channels. However, if during these activities any 
door is determined to be restrained, not fully closed from a 
previous activity, or otherwise not operable, then separate entry 
into Action b is required.
    Thus, based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

[[Page 9389]]

    Because the TS and [TS] Bases changes do not involve any 
physical changes to the ice condenser, or make any changes in the 
operational or maintenance aspects of the ice condenser as required 
by the TSs, there can be no new accidents created from those already 
identified and evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Design basis accident analysis have shown that with 85 percent 
of the total flow area available (uniformly distributed), the ice 
condenser will perform its intended function. Thus, the safety limit 
for ice condenser operability is a maximum 15 percent blockage of 
flow channels. Surveillance Requirement (SR) 4.6.5.1 currently 
applies the 15 percent flow blockage criteria to the total flow area 
of each bay which includes flow passages between the ice baskets, 
past lattice frames, through intermediate and top deck floor 
grating, or past lower plenum support structures and turning vanes. 
This application of the criteria does not have direct correlation to 
the safety limit for blockage of ice condenser flow channels (those 
areas that comprise the area between ice baskets, and past lattice 
frames and wall panels). Changing the TS to implement a surveillance 
program that uses acceptance criteria consistent with the transient 
mass distribution (TMD) analysis will not reduce the margin of 
safety.
    Additionally, verifying an ice bed is left with less than or 
equal to 15 percent flow channel blockage at the end of a refueling 
outage assures the ice bed will remain in an acceptable condition 
for the duration of the operating cycle. During the operating cycle, 
a certain amount of ice sublimates and reforms as frost on the 
colder surfaces in the ice condenser. However, frost has been 
determined to not degrade the flow channel flow area. Thus, design 
limits for the continued safe function of containment subcompartment 
walls and the steel containment vessel are not exceeded due to this 
change.
    The change made to TS 3.6.5.3 Bases does not affect the margin 
of safety as defined in any TS as it does not involve design 
specifications or acceptance criteria. This change only adds a 
clarifying note that entry into Action b is not required solely 
because of actions (standing on and opening intermediate/upper deck 
doors) necessary for the performance of required ice condenser 
surveillances, maintenance, or routine activities. This does not 
preclude entry into Action b during performance of these activities 
should an intermediate deck door or upper deck door otherwise be 
determined inoperable.

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

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 20852. 
Publicly available records will be accessible electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

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

    Date of application for amendment: February 28, 2000, as 
supplemented by letters dated May 12, May 24, June 1 and June 28, 2000.
    Brief description of amendment: The amendments revise certain 
license conditions to reflect the change in ownership interest from 
PECO to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 137.
    Facility Operating License No. NPF-62: The amendment revised the 
License.
    Date of initial notice in Federal Register: April 11, 2000 ( 65 FR 
19396). The May 12, May 24, June 1, and June 28, 2000, supplemental 
letters provided additional clarifying information and did not change 
the staff's original no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 6, 2000 (U-603332).
    Brief description of amendment: The amendment removes from the 
Technical Specification surveillance requirements the minimum operating 
time specified for the containment/drywell hydrogen mixing system.
    Date of issuance: January 25, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 138.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71132) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 25, 2001.
    No significant hazards consideration comments recieved: No.

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

    Date of application for amendment: December 1, 1999, as 
supplemented on September 15, 2000.
    Brief description of amendment: The proposed amendment revised the 
Technical Specifications to change the standard by which you test 
charcoal used in engineered safeguards features systems to American 
Society for Testing and Materials D3803-1989. These revisions are made 
in accordance with Generic Letter 99-02.
    Date of Issuance: January 24, 2001.

[[Page 9390]]

    Effective date: January 24, 2001 and shall be implemented within 30 
days of issuance.
    Amendment No.: 219.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 17, 2000 (65 FR 
31357)
    The September 15, 2000, letter provided clarifying information 
within the scope of the original application and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 24, 2001.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

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

    Date of application for amendments: December 20, 1999, as 
supplemented on January 14, March 10, March 23, March 29, and June 16, 
2000.
    Brief description of amendments: The amendments revise the licenses 
and technical specifications to reflect the transfer of the licenses 
from Commonwealth Edison Company to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: Immediately to be implemented within 30 days.
    Amendment Nos.: 109 & 115.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12583) and (65 FR 12584). The March 10, March 23, March 29, and June 
16, 2000, supplemental letters provided additional clarifying 
information and did not change the staff's original no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
August 3, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-10, 50-237, and 50-249, 
Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County, 
Illinois

    Date of application for amendments: December 20, 1999, as 
supplemented January 14, March 10, March 23, March 29, and June 16 
2000.
    Brief description of amendments: The amendments revise the licenses 
to reflect the transfer of the licenses from Commonwealth Edison 
Company to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 40, 183, and 178.
    Facility Operating License Nos. DPR-2, DPR-19 and DPR-25: The 
amendments revised the Licenses to reflect the transfer of the licenses 
from Commonwealth Edison Company to Exelon Generation Company, LLC.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12582).
    The March 10, March 23, March 29 and June 16, 2000 letters are 
within the scope of the original notice and did not change the original 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: December 20, 1999, as 
supplemented January 14, March 10, March 23, March 29, and June 16, 
2000.
    Brief description of amendments: The amendments revise the licenses 
and technical specifications to reflect the transfer of the license 
from Commonwealth Edison Company to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: Immediately to be implemented within 30 days.
    Amendment Nos.: 146 and 132.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12585). The March 10, March 23, March 29, and June 16, 2000, 
supplemental letters provided additional clarifying information and did 
not change the staff's original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: December 20, 1999, as 
supplemented January 14, March 10, March 23, March 29, and June 16, 
2000.
    Brief description of amendments: The amendments revise the license 
to reflect the transfer of the license from Commonwealth Edison Company 
to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: January 12, 2001.
    Amendment Nos.: 197 and 193.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Licenses.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12581). The March 10, March 23, March 29, and June 16, 2000, 
supplemental letters provided additional clarifying information and did 
not change the staff's original no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois

    Date of application for amendments: December 20, 1999, as 
supplemented January 14, March 10, March 23, March 29, and June 16, 
2000.
    Brief description of amendments: The amendments revised the 
operating licenses to reflect the transfer of the licenses from 
Commonwealth Edison Company to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: January 12, 2001, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: 181 and 168.
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12586).
    The March 10, March 23, March 29, and June 16, 2000, supplemental 
letters provided additional clarifying information and did not change 
the staff's original no significant hazard consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No.

[[Page 9391]]

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: November 17, 1999, as 
supplemented June 14, November 13 and December 4, 2000.
    Brief description of amendment: Increased the allowed outage time 
to restore an inoperable emergency diesel generator set to operable 
status from 72 hours to 14 days.
    Date of Issuance: January 19, 2001.
    Effective Date: January 19, 2001.
    Amendment No.: 170.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (65 
FR 70089). The June 14, November 13, and December 4, 2000, supplements 
did not affect the original proposed no significant hazards 
determination, or expand the scope of the request as noticed in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 19, 2001.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: July 19, 2000.
    Brief description of amendment: Revised the license: (1) to 
implement Siemens Power Corporation (SPC) high thermal performance fuel 
assembly design in Cycle 17, (2) relocate shutdown margin requirements 
in Modes 1 to 5 to the Core Operating Limits Report (COLR), (3) update 
the COLR methodologies listed in the Technical Specification (TS) 
Section 6.9.1.11, and (4) request relief from the SPC fuel assembly 
reconstitution restrictions for peripheral low power fuel assemblies. 
Additionally, administrative changes were made to the boron 
concentration specifications related to the boration requirements.
    Date of Issuance: January 25, 2001.
    Effective Date: January 25, 2001.
    Amendment No.: 171.
    Facility Operating License No. DPR-67: Amendment revised the TSs.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48748).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 25, 2001.
    No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: April 19, 2000.
    Brief description of amendment: The amendment changes Technical 
Specifications (TS) 3.8.4.1, ``Electrical Power System--Containment 
Penetration Conductor Overcurrent Protective Devices;'' 3.8.4.2.1, 
``Electrical Power Systems--Motor-Operated Valves Thermal Overload 
Protections;'' and 3.8.4.2.2, ``Electrical Power Systems--Motor-
Operated Valves Thermal Overload Protection Not Bypassed.'' The 
proposed changes would relocate the requirements for containment 
penetration conductor overcurrent and motor-operated valve thermal 
overload protective devices from the TS to the licensee's Technical 
Requirements Manual (TRM). The Bases for these TSs would also be 
relocated to the TRM.
    Date of issuance: January 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 192.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51360).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 19, 2000.
    Brief description of amendment: The amendment revises the Standby 
Liquid Control boron solution requirements in TS Figure 3.1.7-1 to 
ensure a minimum boron concentration of 660 parts per million in the 
reactor.
    Date of issuance: January 23, 2001.
    Effective date: As of the date of issuance and shall be implemented 
before entering Mode 2 during Cycle 18.
    Amendment No.: 236.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 
65343).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 23, 2001.
    No significant hazards consideration comments received: No.

PECO Energy Company, Docket No. 50-353, Limerick Generating Station, 
Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: October 14, 1999, as 
supplemented February 11, September 22, and October 18, 2000.
    Brief description of amendment: This amendment revised TS Section 
2.2, ``Safety Limits and Limiting Safety Systems Settings,'' and TS 
Section 3.0/4.0, ``Limiting Conditions for Operation and Surveillance 
Requirements.'' These revisions will support the installation of LGS 
Modification P00224 for Unit 2, which will install a Power Range 
Neutron Monitoring System and incorporate long-term thermal-hydraulic 
stability solution hardware.
    Date of issuance: January 16, 2001.
    Effective date: As of date of issuance and shall be implemented 
during the Limerick Unit 2 refueling outage scheduled to begin in the 
spring of 2001.
    Amendment No.: 109.
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67337). The February 11, September 22, and October 18, 2000, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the scope of the original Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 2001.
    No significant hazards consideration comments received: No.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating 
Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: December 20, 1999, as 
supplemented January 3, February 14, March 10, March 23, March 30, and 
June 15, 2000.
    Brief description of amendments: The amendments revised the 
licenses for Limerick Units 1 and 2 to reflect the transfer of PECO's 
ownership of these units to Exelon Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 147 and 108.

[[Page 9392]]

    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the license.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12587).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: December 20, 1999, as 
supplemented December 22, 1999, January 3, February 14, March 10, March 
23, March 30, and June 15, 2000.
    Brief description of amendments: The amendments revise the licenses 
to reflect the transfer of PECO Energy Company's ownership interest in 
the Salem Nuclear Generating Station, Unit Nos. 1 and 2, to Exelon 
Generation Company, LLC.
    Date of issuance: January 12, 2001.
    Effective date: January 12, 2001.
    Amendment Nos.: 241 & 222.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12591). The December 22, 1999, January 3, February 14, March 10, March 
23, March 30, and June 15, 2000, supplements did not expand the scope 
of the original application with respect to both the proposed transfer 
action and the proposed amendment action as initially noticed in the 
Federal Register. No hearing requests or comments were received. In 
addition, the submittal did not affect the applicability of the 
Commission's generic no significant hazards consideration determination 
set forth in 10 CFR 2.1315.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 30, 2000.
    Brief description of amendments: The amendments revised main steam 
isolation valve surveillance testing requirements. Specifically, the 
amendments permit use of the minimum pathway leakage value for the 
``as-found'' test limit.
    Date of issuance: January 24, 2001.
    Effective date: January 24, 2001.
    Amendment Nos.: 267 and 227.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 24, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 7, 2000, as supplemented 
June 23, August 24, September 26, October 6, October 27 and November 
16, 2000.
    Brief description of amendment: The amendment changes the Facility 
Operating License (FOL) and the Technical Specifications (TS) to 
reflect an increase in the full core power rating from 3411 to 3459 
megawatts thermal.
    Date of issuance: January 19, 2001.
    Effective date: January 19, 2001.
    Amendment No.: 31.
    Facility Operating License No. NPF-90: Amendment revises the FOL 
and TS.
    Date of initial notice in Federal Register: September 7, 2000 (65 
FR 54322).
    The Commission's related evaluation of the amendment is contained 
in an Environmental Assessment dated November 21, 2000 and in a Safety 
Evaluation dated January 19, 2001.
    No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: October 25, 2000.
    Brief description of amendment: The amendment makes editorial and 
administrative changes to the Technical Specifications (TSs). These 
changes correct spelling and grammatical errors, correct references, 
eliminate excessive detail related to specifying a job title, revise 
position titles, consolidate pages and generalize statements allowing 
U.S. Nuclear Regulatory Commission (NRC) approved alternatives to 
specified requirements.
    Date of Issuance: January 23, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 196.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71140).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 23, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 31st day of January 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-3028 Filed 2-6-01; 8:45 am]
BILLING CODE 7590-01-P