[Federal Register Volume 66, Number 133 (Wednesday, July 11, 2001)]
[Notices]
[Pages 36335-36347]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-17223]


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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 June 18, 2001 through June 29, 2001. The 
last biweekly notice was published on June 27, 2001.

[[Page 36336]]

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

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

[[Page 36337]]

Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Branch, or may be delivered to the Commission's Public Document Room, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland 20852, by the above date. A copy of the petition 
should also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and to the attorney 
for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Assess and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-
415-4737 or by email to [email protected].

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

    Date of amendment request: May 31, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to incorporate Cycle 14 specific 
limits for the variable low reactor coolant system pressure-temperature 
core protection safety limits. The proposed limits are developed in 
accordance with the methods described in NRC (Nuclear Regulatory 
Commission)-approved Topical Report BAW-10179P-A, ``Safety Criteria and 
Methodology for Acceptable Cycle Reload Analyses.''
    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 Technical Specification limits (Figure 2.1-1) are 
developed in accordance with the methods and assumptions described 
in NRC-approved Framatome ANP Topical Report BAW-10179P-A, ``Safety 
Criteria and Methodology for Acceptable Cycle Reload Analyses.''
    These limits remain bounded by the existing reactor protection 
system (RPS) trip setpoints. The TMI Unit 1 Cycle 14 core introduces 
the Framatome ANP Mark-B12 fuel design. The Mark-B12 fuel design is 
mechanically and hydraulically similar to fuel designs currently in 
use at TMI Unit 1. While the designs are hydraulically similar, the 
Mark-B12 contains a fine mesh debris filter that alters the flow 
characteristics at the core inlet relative to the resident fuel 
designs resulting in the identification of a transition core DNB 
[departure from nucleate boiling] penalty. The higher minimum RCS 
flow requirement (105.5%) applied to offset the transition core DNB 
penalty is bounded by the minimum RCS flow assumed in current 
Updated Final Safety Analyis Report (UFSAR) Chapter 14 accident 
analyses (102%).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed Technical Specification limits (Figure 2.1-1) 
provide core protection safety limits developed in accordance with 
NRC-approved methods and assumptions. The revised Technical 
Specifications limits remain bounded by the existing reactor 
protection trip setpoints. The TMI Unit 1 Cycle 14 core introduces 
the Framatome ANP Mark-B12 fuel design. The Mark-B12 fuel design is 
mechanically and hydraulically similar to the fuel designs currently 
in use at TMI Unit 1. While the designs are hydraulically similar, 
the Mark-B12 contains a fine mesh debris filter that alters the flow 
characteristics at the core inlet relative to the resident fuel 
designs resulting in the identification of a transition core DNB 
penalty. The higher minimum reactor coolant system flow required for 
the transition cycles (105.5%) is within the current range of 
allowable operating flow rates since this value exceeds the minimum 
flow rate assumed for Chapter 14 accident analyses (102%) and is 
well below the maximum flow limit for fuel assembly lift which is 
typically approximately 115% of design flow (depending on fuel type 
and 4th reactor coolant pump startup temperature).
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The existing RPS reactor coolant pressure and temperature trip 
setpoints bound the proposed Technical Specification core protection 
safety limits. The proposed safety limits are developed in 
accordance with NRC-approved safety methods and assumptions. The 
higher minimum reactor coolant system flow requirement assures safe 
operation commensurate with the introduction of the Mark-B12 fuel 
design into the TMI Unit 1 core.
    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: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: March 2, 2001.
    Description of amendment request: The proposed changes would modify 
Technical Specification (TS) 3.3.2, ``Instrumentation--Engineered 
Safety Features Actuation System Instrumentation.'' The Bases of the 
affected TS will also be modified to reflect this change. The proposed 
changes will extend the required surveillance interval for Potter & 
Brumfield MDR Series slave relays, which are installed in the Millstone 
Unit No. 3 Engineered Safety Features Actuation System, from a 
quarterly surveillance interval to an 18-month frequency surveillance 
interval for those relays that meet the reliability assessment criteria 
established by Westinghouse Electric Corporation.
    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 review is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This change to the Technical Specifications does not result in a 
condition

[[Page 36338]]

where the design, material, and construction standards that were 
applicable prior to the change are altered. The same ESFAS 
instrumentation is being used and the same ESFAS system reliability 
is expected. The proposed change will not modify any system 
interface and could not increase the likelihood of an accident since 
these events are independent of this change. The proposed activity 
will not change, degrade or prevent actions or alter assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the SAR [Safety Analysis Report]. Therefore, 
the proposed amendment does not result in any increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This change does not alter the performance of the ESFAS 
[engineered safety features actuation system] mitigation systems 
assumed in the plant safety analysis. Changing the surveillance 
interval for periodically verifying ESFAS slave relays (assuring 
equipment operability) will not create any new accident initiators 
or scenarios. Implementation of the proposed amendment does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.

    This change does not affect the total ESFAS system response assumed 
in the safety analysis. The periodic slave relay functional 
verification is relaxed because of the demonstrated high reliability of 
the relay and its insensitivity to any short term wear or aging 
effects. It is thus concluded that the proposed license amendment 
request does not result in a reduction in margin with respect to plant 
safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06385.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 23, 2001.
    Description of amendment request: The proposed changes would modify 
Technical Specification (TS) 3.1.2.1, ``Reactivity Control Systems--
Boration Systems--Flow Path--Shutdown;'' 3.1.2.2, ``Reactivity Control 
Systems--Flow Paths--Operating;'' 3.1.2.3, ``Reactivity Control 
Systems--Charging Pump--Shutdown;'' 3.1.2.4, ``Reactivity Control 
Systems--Charging Pumps--Operating;'' 3.1.2.5, ``Reactivity Control 
Systems--Borated Water Source--Shutdown;'' 3.1.2.6, ``Reactivity 
Control Systems--Borated Water Sources--Operating;'' 3.4.1.2, ``Reactor 
Coolant System--Hot Standby;'' 3.4.1.3, ``Reactor Coolant System--Hot 
Shutdown;'' 3.4.1.4.1, ``Reactor Coolant System--Cold Shutdown--Loops 
Filled;'' 3.4.1.4.2, ``Reactor Coolant System--Cold Shutdown--Loops Not 
Filled;'' 3.4.1.6, ``Reactor Coolant System--Isolated Loop Startup;'' 
3.4.2.1, ``Reactor Coolant System--Safety Valves--Shutdown;'' 3.4.2.2, 
``Reactor Coolant System--Operating;'' 3.4.9.1, ``Reactor Coolant 
System--Pressure/Temperature Limits;'' and 3.4.9.3, ``Reactor Coolant 
System--Overpressure Protection Systems.'' The Index and the associated 
Bases for these Technical Specifications will be modified as a result 
of the proposed changes.
    The above proposed TS changes will relocate the boration subsystem 
and Residual Heat Removal System over-pressurization protection 
requirements to a licensee-controlled document; modify the Reactor 
Coolant System (RCS) pressure/temperature limits; modify the Cold 
Overpressure Protection System (COPPS) set-point curves, COPPS enable 
temperatures and associated restrictions; modify the reactor vessel 
material surveillance withdrawal schedule; modify the pressurizer code 
safety valve requirements; modify the isolated RCS loop startup 
requirements; and provide numerous minor enhancements to the current 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's analysis, which is 
based on the representations made by the licensee in the April 23, 
2001, application, is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS changes associated with the relocation of the 
boration subsystem requirements to a licensee-controlled document 
and with the revised reactor vessel analyses will not cause an 
accident to occur and will not result in any change in the operation 
of the associated accident mitigation equipment. Therefore, the 
proposed changes will not increase the probability or consequences 
of an accident previously evaluated.
    The proposed TS changes associated with the relocation of the 
Mode 4 and Mode 5 plant restrictions associated with protection of 
the Residual Heat Removal System to a licensee-controlled document 
do not change the design-basis accidents of the same postulated 
events described in the Millstone Unit No. 3 Final Safety Analysis 
Report (FSAR). Therefore, the proposed changes will not increase the 
probability or consequences of an accident previously evaluated.
    The proposed TS changes associated with the pressurizer code 
safety valves do not change the design-basis accidents described in 
the Millstone Unit No. 3 FSAR. Therefore, the proposed changes will 
not increase the probability or consequences of an accident 
previously evaluated.
    The proposed TS changes to the Modes 5 and 6 restrictions 
associated with restoration of an isolated RCS loop do not change 
the design-basis accidents of the same postulated events described 
in the Millstone Unit No. 3 FSAR. Therefore, the proposed changes 
will not increase the probability or consequences of an accident 
previously evaluated.
    The additional proposed changes to the TS that will standardize 
terminology, relocate information to the Bases, remove extraneous 
information, modify the requirements to prevent rod withdrawal for 
operational flexibility, and make minor format changes will not 
result in any technical changes to the current requirements. 
Therefore, these additional proposed changes will not increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the TSs do not impact any system or 
component that could cause an accident, nor will it alter the plant 
configuration or require any unusual operator actions, nor will it 
alter the way any structure, system, or component functions. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.

    The revised analyses are based on American Society of Mechanical 
Engineers (ASME) Section XI Code Case N-640, which provides an 
alternate reference fracture toughness curve (KIc) for 
establishment of the beltline P/T limits. The analyses restrictions are 
less restrictive than those associated with the current analyses. 
However, the reduction in the margin of safety is small relative to the 
conservatism provided by ASME Section XI margins. Therefore, the 
proposed changes will not result in a significant reduction in a margin 
of safety.
    The proposed TS changes associated with the relocation of the 
boration subsystem and RHR System overpressure protection requirements 
to a licensee-controlled document, pressurizer code safety valve

[[Page 36339]]

requirements, and isolated RCS loop startup will not result in a 
significant reduction in a margin of safety.
    The additional proposed changes to the TSs that will standardize 
terminology, relocate information to the Bases, remove extraneous 
information, modify requirements to prevent rod withdrawal for 
operational flexibility, and make minor format changes will not result 
in any technical changes to the current requirements. Therefore, these 
additional changes will not result in a significant reduction in a 
margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06385.
    NRC Section Chief: James W. Clifford.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina, Docket Nos. 
50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, 
South Carolina

    Date of amendment request: March 22, 2001.
    Description of amendment request: The proposed amendments would 
revise the Catawba Nuclear Station (CNS) Unit 1 and Unit 2, and the 
McGuire Nuclear Station (MNS) Unit 1 and Unit 2 Technical 
Specifications (TS). The proposed TS revisions are presented in two 
parts. Part I affects the current MNS and CNS TS surveillance 
requirement (SR), and associated TS Bases for the methodology and 
frequency for the chemical analyses of the ice condenser ice bed 
(stored ice). The revision results in renumbering the SRs. Also, this 
proposed amendment adds a new TS SR to address sampling requirements 
for ice additions to the ice bed. Part II proposes revisions to the 
current MNS and CNS TS surveillance requirement (SR) acceptance 
criteria and surveillance frequency for the inspection of ice condenser 
ice basket flow channel areas. The proposed change also results in 
renumbering the SRs. Associated changes are also made to the TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The Proposed Change 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. The proposed changes to the TSs and associated TS Bases are 
solely to revise and provide clarification of the ice sampling and 
chemical analysis requirements, and flow area verification 
requirements. Since these proposed changes do not result in, or 
require, any physical change to the ice condenser, then there can be 
no change in the probability of an accident previously evaluated in 
the SAR.
    In order for the consequences of any previously evaluated event 
to be changed, there would have to be a change in the ice 
condenser's physical operation during a LOCA or HELB, or in the 
chemical composition of the stored ice.
    The proposed changes add an upper limit on boron concentration, 
which is the bounding value for the boron precipitation analysis. 
The upper limit boron concentration is an existing DBA analysis 
input limit that is controlled by existing procedure. Therefore, the 
addition of a TS requirement for an upper limit on boron 
concentration does not affect the physical operation or condition of 
the ice condenser.
    Though the frequency of the existing surveillance requirement 
for sampling the stored ice is changed from once every 18 months to 
once every 54 months, the sampling requirements are strengthened 
overall with the requirement to obtain one randomly selected sample 
from each ice condenser bay (24 total samples) rather than nine 
``representative'' samples, and the addition of a new surveillance 
requirement to verify each addition of ice meets the existing 
requirements for boron concentration and pH value.
    The proposed changes clarify that each sample of stored ice is 
individually analyzed for boron concentration and pH, and that the 
acceptance criteria for each parameter is based on the average 
values obtained for the 24 samples. This is consistent with the 
bases for the boron concentration of the ice, which is to ensure the 
accident analysis assumptions for containment sump pH and boron 
concentration are not altered following complete melting of the ice 
condenser. Historically, chemical analysis of the stored ice has had 
a very limited number of instances where an individual sample did 
not meet the boron or pH requirements, with all subsequent 
evaluations (follow up sampling) showing the ice condenser as a 
whole was well within these requirements. Requiring chemical 
analysis of each sample is provided to preclude the practice of 
melting all samples together before performing the analysis, and to 
ensure the licensee is alerted to any localized anomalies for 
investigation and resolution without the burden of entering a 24 
hour ACTION Condition, provided the averaged results are acceptable.
    The proposed changes revise and clarify the flow area 
verification requirements. Regarding the consequences of analyzed 
accidents, the ice condenser is an engineered safety feature 
designed, in part, to limit the containment subcompartment and steel 
vessel pressures immediately following the initiation of a LOCA or 
HELB. Conservative sub-compartment pressure analysis shows this 
criteria will be met if the reduction in the flow area per bay 
provided for ice condenser air/steam flow paths is 15 
percent, or if the total flow area blocked within each lumped 
analysis section is 15 percent as assumed in the safety 
analysis. The present 0.38 inch frost/ice buildup surveillance 
criteria only addresses the acceptability of any given flow path, 
and has no existing correlation between flow paths exceeding this 
criteria and percent of total flow path blockage. In fact, it was 
never the intent of the current surveillance requirement (SR) to 
make such a correlation. If problems were encountered in meeting the 
0.38 inch criteria, it was expected that additional inspection and 
analysis, such as provided in the proposed amendment, would be 
performed to make such a determination. Thus, the proposed amendment 
for flow blockage determination provides the necessary assurance 
that flow path requirements are met without additional evaluations.
    The proposed amendment also revises the flow area verification 
surveillance frequency from every 9 months to every 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. By verifying an ice bed condition of less than or 
equal to 15% flow channel blockage following completion of these 
maintenance activities, the surveillance assures the ice bed is in 
an acceptable condition for the duration of the operating cycle. 
During the operating cycle, an expected amount of ice sublimates and 
reforms as frost on the colder surfaces in the Ice Condenser. 
However, frost does not degrade flow channel flow area per the 
Westinghouse definition of frost. The surveillance will effectively 
demonstrate operability for an allowed 18 month cycle. Therefore, 
increasing the surveillance frequency does not affect the ice 
condenser operation or accident response. An ice bed condition of 
less than or equal to 15% flow channel blockage is assured to be 
maintained for the operating cycle to address the limiting design 
basis accident(s) (DBAs).
    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 Change Does Not Create The Possibility Of A New 
Or Different Kind Of Accident From Any Accident Previously 
Evaluated[.]

[[Page 36340]]

    Because the TSs and TS Bases changes do not involve any physical 
changes to the ice condenser, any physical or chemical changes to 
the ice contained therein, 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 Change Does Not Involve A Significant Reduction 
In A Margin Of Safety[.]
    The ice condenser Technical Specifications ensure that during a 
LOCA or HELB the ice condenser will initially pass sufficient air 
and steam mass to preclude over pressurizing lower containment, that 
it will absorb sufficient heat energy initially and over a 
prescribed time period to assist in precluding containment vessel 
failure, and that it will not alter the bulk containment sump pH and 
boron concentration assumed in the accident analyses.
    Since the proposed changes do not physically alter the ice 
condenser, but rather only serve to strengthen and clarify ice 
sampling and analysis requirements, the only area of potential 
concern is the effect these changes could have on bulk containment 
sump pH and boron concentration following ice melt. However, this is 
not affected because there is no change in the existing requirements 
for pH and boron concentration, except to add an upper limit on 
boron concentration. This upper limit is the bounding value for the 
boron precipitation analysis. The upper limit boron concentration is 
an existing design bases limit that is controlled by existing 
procedure. Therefore, the addition of a TS requirement for an upper 
limit on boron concentration does not affect the physical operation 
or condition of the ice condenser.
    Averaging the pH and boron values obtained from analysis of the 
individual samples taken is not a new practice, just one that was 
not consistently used by all ice condenser plants. Using the 
averaged values provides an equivalent bulk value for the ice 
condenser, which is consistent with the accident analysis for the 
bulk pH and boron concentration of the containment sump following 
ice melt.
    Changing the performance Frequency for sampling the stored ice 
does not reduce any margin of safety because (1) the newly proposed 
surveillance ensures ice additions meet the existing boron 
concentration and pH requirements, (2) there are no normal operating 
mechanisms, including sublimation, that reduce the ice condenser 
bulk pH and boron concentration, and (3) the number of required 
samples has been increased from 9 to 24 (one randomly selected ice 
basket per bay), which is approximately the same number of samples 
that would have been taken in the same time period under the 
existing requirements.
    Design Basis Accident analyses 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. The existing TS surveillance requirement currently 
uses a specific value of 0.38 inch buildup to determine if 
unacceptable frost/ice blockage exists in the ice condenser. 
However, this specific value does not have a direct correlation to 
the safety limit for blockage of ice condenser flow area. The 
proposed TS amendment requires more extensive visual inspection (33 
percent of the flow area/bay) than is currently described (2 flow 
channels/bay) in the TS Bases, thus providing greater reliability 
and a direct relationship to the analytical safety limits. Changing 
the TS to implement a surveillance program that is more reliable and 
uses acceptance criteria of less than or equal to 15 percent flow 
blockage, as allowed by the TMD code analysis, will not reduce the 
margin of safety of any TS.
    The proposed amendment also revises the surveillance frequency 
of flow area inspection from every 9 months to every 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. By verifying an ice bed condition of less than or 
equal to 15% flow channel blockage following completion of these 
maintenance activities, the surveillance assures the ice bed is in 
an acceptable condition for the duration of the operating cycle. 
During the operating cycle, an expected amount of ice sublimates and 
reforms as frost on the colder surfaces in the Ice Condenser. 
However, frost has been determined to not degrade flow channel flow 
area. Thus, design limits for the continued safe function of 
containment sub-compartment walls and the steel containment vessel 
are not exceeded due to this change.
    Thus, it can be concluded that the proposed TS and TS Bases 
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: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Richard Emch.

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

    Date of amendment request: June 12, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.8.1.1.2.a and delete the table 
referenced in TS 4.8.1.1.2.a, to remove the requirement for an 
accelerated test frequency for the emergency diesel generators (EDGs); 
delete TS 4.8.1.1.2.c.1 to remove the requirement to subject the EDG to 
an inspection in accordance with the manufacturer's recommendations; 
revise TSs 4.8.1.1.2.c.9, 10 and 13 to allow that EDG surveillances 
regarding the 24-hour endurance run, the auto-connected loads not 
exceeding the 2-hour rating, and the fuel transfer pump transferring 
fuel via the cross-connect lines, are conducted during modes other than 
during shutdown; and delete TSs 4.8.1.1.3 and 6.9.1.5.d to remove the 
EDG special reporting requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    There are no previously evaluated accidents associated with 
these surveillance activities. The EDGs are not accident initiators. 
The EDGs provide assistance in accident mitigation. There are no 
technical changes related to the acceptance criteria of any of these 
surveillances nor are there any physical changes to plant design 
proposed in this amendment request. The proposed change, requesting 
that the frequency and scheduling aspects of the surveillance 
requirements be changed to accommodate improved planning capability 
for testing and maintenance activities, does not affect the accident 
analyses. Additionally, the allowance to perform testing and 
maintenance activities on line will improve EDG availability during 
periods of shutdown operations.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The proposed change does not include any physical changes to 
plant design or a change to any of the current SR [surveillance 
requirement] acceptance criteria. Performance of any of these 
surveillance activities while at power does not render the EDGs 
unavailable in that they can provide station power on demand. 
Performance of maintenance activities and surveillance requirements 
while on line, which could result in the equipment being out of 
service, was included in the development of the Limiting Conditions 
for Operation (LCO). Quantitative and qualitative evaluations 
relative to the credit allowed for redundant components and the time 
allowed for corrective actions were also considered in LCO 
development. Performance of these activities while on line does not 
create any new or different kinds of accident. The capability of the 
EDG to respond to an accident situation while tied to the grid

[[Page 36341]]

during testing activities is tested as required by existing 
surveillance requirements. These tests ensure that if tied to the 
gird [grid] the EDG output breaker will open and the EDG [will] 
remain running in standby until an under voltage condition is 
observed, at which time the EDG will automatically tie on to the 
4160 V [volt] ESF [engineered safety features] bus.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The proposed changes are associated with surveillance 
requirements for the EDGs. The deletion of accelerated testing 
requirements provides an enhancement to safety by eliminating 
unnecessary testing. The remaining proposed changes allow certain 
EDG surveillance requirements to be performed when the plant is at 
power rather than when shutdown. The operation of, and requirements 
for, the equipment covered by the affected TSs will remain 
essentially the same.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: June 12, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.8.1.1 to provide a one-time 
extension of the allowed outage time (AOT) for an inoperable emergency 
diesel generator (EDG) from three days to ten days, provided the 
alternate alternating-current diesel generator (AACDG) is available. In 
addition, the proposed amendment would revise TS 3.4.4 to make the 
action associated with an inoperable emergency power supply to the 
pressurizer heaters consistent with the proposed EDG AOT.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Typically, only one EDG is OOS [out of service] for maintenance 
activities at any given time. The standby EDG is aligned as required 
by Technical Specifications (TS) and available for auto start upon 
demand. Additionally, the AACDG is verified available and capable of 
being aligned to the Engineered Safety Features (ESF) electrical 
buses associated with the OOS EDG. The AACDG is sized such that it 
can carry the loads equivalent to at least one train of Engineered 
Safety Features (ESF) equipment. In the event of a loss of offsite 
power while an EDG is OOS, the AACDG will be manually started and 
loaded. The time delay associated with the manual start of the AACDG 
will result in a minimal change in the overall risk associated with 
the ability to reestablish power to ESF equipment upon a loss of 
offsite power. However, assuming the standby EDG operates as 
designed, it will start upon receipt of the automatic start signal 
and sequence on loads as required.
    The plant can be maintained in a safe configuration or mitigate 
any accident situations with only one train of ESF components. 
Reliance upon the AACDG to provide a backup function ensures a 
minimal change in risk associated with extending the EDG AOT. The 
EDG AOT of 72 hours under the existing technical specifications does 
not consider an additional backup power supply to be available to 
mitigate a loss of offsite power. The proposed change will ensure 
that an alternate onsite diesel generator will be available while 
the EDG is out of service. Therefore, this change is considered a 
more responsive action than that contained in the current TSs.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The duration of an AOT is determined considering that there is a 
minimal possibility that an accident will occur while a component is 
removed from service. Typically, only the single redundant train is 
available during the AOT with no backup components available to 
supply the function of the component. The proposed change allows the 
EDG AOT to be extended one time for each EDG to 10 days with 
reliance on the AACDG. If the AACDG is not available, the AOT is 72 
hours. No new modifications are required to allow the AACDG to 
function.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The EDG AOT will be 10 days for each EDG if the AACDG is 
available. However, if the AACDG is not available, the EDG AOT will 
remain at 72 hours. The AACDG supplies backup power to the redundant 
train of ESF components. The standby EDG, which is typically not 
aligned in a test mode during the AOT, will be available to 
automatically start and sequence on loads upon demand. Two trains of 
ESF components powered from the onsite electrical sources, the 
standby EDG and the AACDG, will be available in the event of an 
accident. When the AACDG is not available, the current 72-hour AOT 
will begin. In conclusion, either the AACDG will be available, which 
will result in two ESF trains being available in the unlikely event 
of an accident, or the current AOT will apply.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

Exelon Generation Company, LLC, PSEG Nuclear LLC, and Atlantic City 
Electric Company, Docket No. 50-278, Peach Bottom Atomic Power Station, 
Unit No. 3, York County, Pennsylvania

    Date of application for amendment: May 30, 2001.
    Description of amendment request: The proposed amendment would 
allow an extension to the interval for integrated leak rate tests 
(ILRTs) of the reactor containment building. The change involves a one-
time exception to the 10-year frequency of the performance-based 
leakage rate testing program for Type A tests as required by Nuclear 
Energy Institute 94-01, Revision 0, ``Industry Guideline for 
Implementing Performance-Based Option of 10 CFR Part 50, Appendix J.'' 
The current 10-year containment building ILRT for Peach Bottom Atomic 
Power Station (PBAPS), Unit 3, is due in December 2001 and is currently 
scheduled to be performed during Refueling Outage 3R13 in October 2001. 
The proposed exception would allow the next ILRT for PBAPS, Unit 3, to 
be performed within 16 years (December 2007) from the last ILRT as 
opposed to the current 10-year frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 36342]]


    1. The proposed Technical Specification change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to Technical Specification 5.5.12 
(``Primary Containment Leakage Rate Testing Program'') involves a 
one-time extension to the current interval for Type A containment 
testing. The current test interval of ten (10) years would be 
extended on a one-time basis to no longer than sixteen (16) years 
from the last Type A test. The proposed Technical Specification 
change does not involve a physical change to the plant or a change 
in the manner in which the plant is operated or controlled. The 
reactor containment is designed to provide an essentially leak tight 
barrier against the uncontrolled release of radioactivity to the 
environment for postulated accidents. As such the reactor 
containment itself and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident. Therefore, the proposed Technical 
Specification change does not involve a significant increase in the 
probability of an accident previously evaluated.
    The proposed change involves only the extension of the interval 
between Type A containment leakage tests. Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by plant Technical Specifications. Industry 
experience has shown, as documented in NUREG-1493, that Type B and C 
containment leakage tests have identified a very large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is very 
small. PBAPS, Unit 3 ILRT test history supports this conclusion. 
NUREG-1493 concluded, in part, that reducing the frequency of Type A 
containment leak tests to once per twenty (20) years leads to an 
imperceptible increase in risk. The integrity of the reactor 
containment is subject to two types of failure mechanisms which can 
be categorized as (1) activity based and (2) time based. Activity 
based failure mechanisms are defined as degradation due to system 
and/or component modifications or maintenance. Local leak rate test 
requirements and administrative controls such as design change 
control and procedural requirements for system restoration ensure 
that containment integrity is not degraded by plant modifications or 
maintenance activities. The design and construction requirements of 
the reactor containment itself combined with the containment 
inspections performed in accordance with ASME Section XI, the 
Maintenance Rule and licensing commitments related to containment 
coatings serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is detectable only by 
Type A testing. Therefore, the proposed Technical Specification 
change does not involve a significant increase in the consequences 
of an accident previously evaluated.
    2. The proposed Technical Specification change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed revision to the Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. The reactor containment and the testing requirements 
invoked to periodically demonstrate the integrity of the reactor 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident and do not involve the prevention or 
identification of any precursors of an accident. The proposed 
Technical Specification change does not involve a physical change to 
the plant or the manner in which the plant is operated or 
controlled. Therefore, the proposed Technical Specification change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed Technical Specification change does not involve 
a significant reduction in a margin of safety.
    The proposed revision to Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. The proposed Technical Specification change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The specific requirements 
and conditions of the Primary Containment Leakage Rate Testing 
Program, as defined in Technical Specifications, exist to ensure 
that the degree of reactor containment structural integrity and 
leak-tightness that is considered in the plant safety analysis is 
maintained. The overall containment leakage rate limit specified by 
Technical Specifications is maintained. The proposed change involves 
only the extension of the interval between Type A containment 
leakage tests. Type B and C containment leakage tests will continue 
to be performed at the frequency currently required by plant 
Technical Specifications.
    PBAPS, Unit 3 and industry experience strongly supports the 
conclusion that Type B and C testing detects a large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is small. The 
containment inspections performed in accordance with ASME Section 
XI, the Maintenance Rule and the Coatings Program serve to provide a 
high degree of assurance that the containment will not degrade in a 
manner that is detectable only by Type A testing. Additionally, the 
on-line containment monitoring capability that is inherent to 
inerted BWR containments allows for the detection of gross 
containment leakage that may develop during power operation. The 
combination of these factors ensures that the margin of safety that 
is inherent in plant safety analysis is maintained. Therefore, the 
proposed Technical Specification change does not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: June 13, 2001.
    Description of amendment request: The proposed amendment clarifies 
the Kewaunee Nuclear Power Plant Technical Specification 5.3 to permit 
lead-test-assemblies to be used, regardless of clad material, as long 
as the Nuclear Regulatory Commission has generically approved the fuel 
assembly design for use in pressurized water reactors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Changing the technical specification within limits of the 
bounding accident analyses cannot change the probability of an 
accident previously evaluated, nor will it increase radiological 
consequence predicted by the analyses of record. Controlling the use 
of lead-test-assemblies, designs of which were approved by the NRC, 
according to limitations approved by the NRC constrains fuel 
performance within limits bounded by existing design basis accident 
and transient analyses. Thus, nothing in this proposal will cause an 
increase in the probability or consequence of an accident previously 
evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Inclusion in the reactor core of lead-test-assemblies according 
to limitations set by the NRC and of a design approved by the NRC 
ensures that their effect on core performance remains within 
existing design limits. Use of NRC approved fuel assemblies as lead-
test-assemblies is consistent with current plant design bases, does 
not adversely affect any fission product barrier, and does not alter 
the safety function of safety significant systems, structures and 
components or their roles in accident prevention or mitigation. 
Currently licensed design basis accident and transient analyses of 
record bound the effect of lead-test-assemblies. Thus, this proposal 
does not create the possibility of a new or different kind of 
accident.

[[Page 36343]]

    (3) Involve a significant reduction in the margin of safety.
    The proposed change does not alter the manner in which Safety 
Limits, Limiting Safety System Setpoints, or Limiting Conditions for 
Operation are determined. This clarification of TS 5.3 is bounded by 
existing limits on reactor operation. It leaves current limitations 
for use of lead-test-assemblies in place, conforms to plant design 
bases, is consistent with current safety analyses, and limits actual 
plant operation within analyzed and licensed boundaries. Thus, 
changes proposed by this request 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: November 16, 2000.
    Description of amendment request: The proposed change would delete 
a note to Technical Specification (TS) Surveillance Requirement (SR) 
3.6.1.1.1, which permitted a temporary extension to the surveillance 
interval for testing spectacle flanges 2S299A and 2S299B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This change to Technical Specification SR 3.6.1.1.1 is 
administrative in nature. The note is no longer required as the 
condition has been corrected and the SR performed with acceptable 
results. Removal of the note restores the Technical Specification to 
its original condition and therefore, this proposed amendment does 
not involve any increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The note is no longer required as the condition has been 
corrected and the SR performed with acceptable results. Removal of 
the note by this change restores the Technical Specification to its 
original condition and therefore does not create any possibility of 
a new or different kind of accident from those previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The note is no longer required as the condition has been 
corrected and the SR performed with acceptable results. Therefore, 
removal of the note by this change restores the Technical 
Specification to its original condition and does not involve a 
reduction in 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(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Acting Section Chief: Richard Correia.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: March 5, 2001.
    Description of amendment request: The proposed amendment would: (1) 
change the Security Plan provision that a member of the security force 
escort all vehicles, other than designated licensee vehicles, and 
delete the related Security Training and Qualification Plan task; (2) 
change the requirement of the Security Plan that all areas of the 
protected area be illuminated to a minimum of 0.2 footcandle; and (3) 
change the frequency of protected area patrols in the Security Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involving security activities do not reduce 
the ability for the security organization to prevent radiological 
sabotage and therefore do not increase the probability or 
consequences of a radiological release previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve functions of the security 
organization concerning vehicle control, protected area 
illumination, and protected area patrol frequency. Analysis of the 
proposed changes has not indicated nor identified a new or different 
kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Analysis of the proposed changes show that they affect only the 
functions of the Security organization and have no impact upon nor 
cause a significant reduction in margin of safety for plant 
operation. The failure points of key safety parameters are not 
affected.

    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of amendment request: May 21, 2001.
    Description of amendment request: The proposed amendment will 
delete the definitions, the limiting conditions for operation, and the 
surveillance requirements and revise the design features and 
administrative controls to reflect the transfer of all the spent 
nuclear fuel from the 10 CFR Part 50 licensed site to the 10 CFR Part 
72 licensed Independent Spent Fuel Storage Installation (ISFSI) from 
the Rancho Seco Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed changes reflect removing the spent nuclear 
[fuel] from the 10 CFR Part 50 licensed facility and transferring 
the fuel to a 10 CFR Part 72 licensed facility. The design basis 
accidents analyzed in the Rancho Seco Defueled Safety Analysis 
Report (DSAR) include the fuel handling accident and a loss of 
offsite power (LOOP). The fuel handling accident is the worst-case 
design basis accident postulated to occur at Rancho Seco. Both of 
these accidents are based on spent nuclear fuel being stored in the 
spent fuel pool at the 10 CFR Part 50 licensed facility.
    With the removal of the spent nuclear fuel from the 10 CFR Part 
50 licensed facility,

[[Page 36344]]

there are no remaining important to safety systems required to be 
monitored and there are no remaining credible accidents that require 
the actions of a Certified Fuel Handler or Non-Certified Fuel 
Handler to prevent occurrence or mitigate the consequences.
    DSAR Section 14.2 provides a discussion of accidents during 
decommissioning. The DSAR concludes that the consequences of the 
accidents evaluated in NUREG/CR-0130 ``Technology, Safety, and Costs 
of Decommissioning a Reference Pressurized Water Reactor Power 
Station'' bound the potential accidents that could occur during 
decommissioning at Rancho Seco. The proposed Technical Specification 
changes have no impact on decommissioning activities.
    The proposed Technical Specification Section D5.2 precludes the 
storage of spent nuclear fuel at the 10 CFR Part 50 licensed 
facility. The probability or consequences of accidents at the ISFSI 
are evaluated in the ISFSI FSAR [Final Safety Analysis Report] and 
are independent of the 10 CFR Part 50 license.
    Therefore, with all of the spent fuel stored at the Rancho Seco 
ISFSI, the accidents evaluated in the DSAR are no longer relevant, 
and the proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes reflect the reduced operational risks 
within the 10 CFR Part 50 licensed facility after the fuel is 
transferred to the 10 CFR Part 72 licensed ISFSI. The proposed 
changes do not result in physical changes to the 10 CFR Part 50 
facility and the plant conditions for which the design basis 
accidents have been evaluated are no longer applicable.
    No new failure modes are introduced as the result of the 
proposed changes. Therefore, the proposed changes will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    No. As described above, the proposed changes reflect the reduced 
operational risks within the 10 CFR Part 50 licensed facility after 
the fuel is transferred to the ISFSI. The design basis and the 
accident assumptions in the Defueled Safety Analysis Report (DSAR), 
and the Technical Specification Bases are no longer applicable after 
the fuel is permanently removed from the 10 CFR Part 50 licensed 
facility. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's significant hazards 
analysis and, based on this review, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Thomas A. Baxter, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N. Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of amendment request: June 7, 2001.
    Description of amendment request: The proposed amendment will 
delete certain administrative requirements from the Rancho Seco 
Technical Specifications and relocate other administrative requirements 
from the Rancho Seco Technical Specifications to the Rancho Seco 
Quality Manual following the transfer of all the spent nuclear fuel 
from the 10 CFR part 50 licensed site to the 10 CFR Part 72 licensed 
Independent Spent Fuel Storage Installation (ISFSI).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed changes are administrative and involve deleting 
certain administrative requirements from the Technical 
Specifications. Some administrative requirements are no longer 
applicable after permanently transferring the spent nuclear fuel 
from the 10 CFR 50 licensed facility to the 10 CFR 72 licensed 
ISFSI. Other administrative requirements are being relocated to the 
NRC-approved Rancho Seco Quality Manual (RSQM).
    Relocating administrative requirements to the NRC-approved RSQM 
is consistent with the guidance in NRC Administrative Letter 95-06. 
Relocating these administrative requirements will not alter the 
configuration or operation of the facility, and therefore does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    In addition, deleting certain administrative requirements (i.e., 
PRC [Plant Review Committee] and MSRC [Management Safety Review 
Committee]) is based on permanently removing the spent nuclear fuel 
from the 10 CFR Part 50 licensed facility and transferring the fuel 
to a 10 CFR Part 72 licensed facility.
    The design basis accidents analyzed in the Rancho Seco Defueled 
Safety Analysis Report (DSAR) include the fuel handling accident and 
a loss of offsite power. Both of these accidents are based on spent 
nuclear fuel being stored in the spent fuel pool at the 10 CFR Part 
50 licensed facility.
    With all of the spent fuel stored at the Rancho Seco ISFSI, the 
accidents evaluated in the DSAR are no longer relevant. Therefore, 
the proposed license amendment does not involve any increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes are administrative and reflect the 
reduced operational risks within the 10 CFR Part 50 licensed 
facility after the fuel is transferred to the 10 CFR Part 72 
licensed ISFSI. The proposed changes do not result in physical 
changes to the 10 CFR Part 50 facility, and the plant conditions for 
which the design basis accidents have been evaluated are no longer 
applicable.
    No new failure modes are introduced as the result of the 
proposed changes. Therefore, the proposed changes will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    No. As described above, the proposed changes are administrative 
and reflect the reduced operational risks within the 10 CFR Part 50 
licensed facility after the fuel is transferred to the ISFSI. The 
design basis and the accident assumptions in the DSAR and the 
Technical Specification Bases are no longer applicable after the 
fuel is permanently removed from the 10 CFR Part 50 licensed 
facility. Therefore, the proposed changes do not involve any 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's significant hazards 
analysis and, based on this review, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Thomas A. Baxter, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N. Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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: May 24, 2001.
    Description of amendment request: The proposed change would 
relocate Technical Specification 3/4.9.6, ``Refueling Machine'' to the 
Technical Requirements Manual consistent with NUREG-1431, ``Standard 
(Improved) Technical Specifications--Westinghouse Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 36345]]

issue of no significant hazards consideration, which is presented 
below:
    Pursuant to 10 CFR 50.92, it has been determined that this proposed 
amendment involves no significant hazards consideration. This 
determination was made by applying the Nuclear Regulatory Commission 
established standards contained in 10 CFR 50.92. These standards assure 
that operation of South Texas Project in accordance with this request 
consider the following:

    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This request involves an administrative change only. No actual 
plant equipment or accident analyses will be affected by the 
proposed changes. Operability of the refueling machine ensures that 
the equipment used to handle fuel within the reactor vessel has 
sufficient load capacity for handling fuel assemblies and/or control 
rods. Although the refueling machine is designed and has interlocks 
that can prevent damage to the fuel assemblies, the equipment is not 
assumed to function or actuate to mitigate the consequences of a 
design basis accident or transient in the safety analysis. 
Therefore, the proposed amendment does not result in any increase in 
the probability or consequences of an accident previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    This request involves an administrative change only. The 
proposed change does not alter the performance of the refueling 
machine and auxiliary hoist or the manner in which the equipment 
will be operated. The refueling equipment will still be tested 
before placing the equipment into operational service. Changing the 
location of these requirements and surveillances from Technical 
Specifications to the Technical Requirements Manual [TRM] will not 
create any new accident initiators or scenarios. Since the proposed 
changes only allow activities that are presently approved and 
conducted, no possibility exists for a new or different kind of 
accident from those previously evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    This request involves administrative changes only. No actual 
plant equipment or accident analyses will be affected by the 
proposed change. Additionally, the proposed changes will not relax 
any criteria used to establish safety limits, will not relax any 
safety systems settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, the proposed changes 
will not impact the margin of safety.

Conclusion

    Based on the above analysis, STPNOC concludes that the proposed 
amendment to relocate these requirements from Technical 
Specifications to the TRM involve no significant hazards 
consideration under the standards set forth in 10 CFR 50.92(c) and, 
accordingly, a finding of ``no significant hazards consideration'' 
is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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.

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: May 24, 2001.
    Description of amendment request: Revise the Technical 
Specification definition for CORE ALTERATIONS so that moving the 
control rods with the integrated head package would not be a core 
alteration.
    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:
    STPNOC [South Texas Project Nuclear Operating Company] has 
evaluated whether or not a significant hazards consideration is 
involved with the proposed amendment by focusing on the three standards 
set forth in 10CFR50.92, ``Issuance of amendment,'' as discussed below.

    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change in the definition of CORE ALTERATIONS will 
not alter the way STPNOC handles the integrated head package. No new 
accident initiators will be introduced. Consequently, there is no 
significant increase in the probability of an accident previously 
evaluated.
    The evaluation demonstrates that the RCCAs [rod cluster control 
assemblies] have no effect on reactivity when they are withdrawn 
into the integrated head package. The proposed change has no effect 
on assumptions made in any accident previously evaluated. 
Consequently, there are no significant increases in the consequences 
of an accident previously evaluated.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any new processes, 
procedures, or significantly different plant configurations. No new 
reactivity configurations are presented. Consequently, the 
possibility of a new or different kind of accident is not created.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The evaluation shows the RCCAs have no effect on reactivity when 
they are withdrawn into the integrated head package. Moving the 
integrated head package with the RCCAs withdrawn provides the same 
degree of control on reactivity as the original definition. 
Consequently, the proposed change does not involve a significant 
reduction in the margin of safety.

Conclusion

    Based upon the analysis provided herein, the proposed amendments 
will not increase the probability or consequences of an accident 
previously evaluated, create the possibility of a new or different 
kind of accident from any accident previously evaluated, or involve 
a reduction in a margin of safety. Therefore, the proposed 
amendments meet the requirements of 10 CFR 50.92 and do not involve 
a significant hazards consideration.

    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.

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

[[Page 36346]]

with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no 
environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

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

    Date of application for amendments: October 20, 2000, as 
supplemented by letter dated March 12, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 3.7.10, ``Control Room Area Ventilation 
System (CRAVS)'' by eliminating the requirement for the CRAVS high 
chlorine protection function. The amendments also eliminated the 
requirement for the safety related chlorine monitor and the capability 
for automatic isolation of the control room area ventilation system 
when prompted by a signal from the detectors. Revisions to the 
corresponding Bases for TS 3.7.10 have been incorporated.
    Date of issuance: June 28, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance June 28, 2001.
    Amendment Nos.: 191 and 183.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2013). The supplement dated March 12, 2001, provided clarifying 
information that did not change the scope of the October 20, 2000, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 28, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 19, 2001.
    Brief description of amendment: The amendment modifies Technical 
Specification 3.6.5, ``Vacuum Relief Valves,'' Limiting Condition for 
Operation, and extends the allowed outage time from 4 hours to 72 hours 
to restore the vacuum relief line to OPERABLE status. In addition, 
Attachment 1 to the Waterford Steam Electric Station, Unit 3 Operating 
License has been deleted and paragraph 2.C.1 revised to reflect the 
deletion.
    Date of issuance: June 18, 2001.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 171.
    Facility Operating License No. NPF-38: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: 66 FR 27176, dated May 
16, 2001.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 18, 2001.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: November 28, 2000, as 
supplemented June 12, 2001.
    Brief description of amendment: This amendment revises the design 
basis for the post-trip steam line break analysis to allow less than or 
equal to 2% fuel failure.
    Date of Issuance: June 19, 2001.
    Effective Date: June 19, 2001.
    Amendment No.: 116.
    Facility Operating License No. NPF-16: Amendment does not revise 
the operating license or its appendices.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9384). The June 12, 2001, supplement did not affect the original 
proposed no significant hazards determination, or expand the scope of 
the request as noticed in the Federal Register. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated June 19, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 30, 2001.
    Brief description of amendment: The amendment changes two 
requirements in the Operating License regarding the reporting of 
changes to the approved fire protection plan and exceeding the licensed 
steady-state power level.
    Date of issuance: June 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 203
    Facility Operating License No. DPR-20. Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17965). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 26, 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: March 2, 2001.
    Brief description of amendment: The proposed amendment revises 
Technical Specifications (TS) Section 5.6, ``TS Bases Control 
Program,'' to delete the term ``unreviewed safety question'' consistent 
with the recent revision to 10 CFR 50.59. The TS, as amended, would 
continue to incorporate the criteria of 10 CFR 50.59 by reference.
    Date of issuance: June 15, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 32.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17971).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 15, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 2nd day of July 2001.


[[Page 36347]]


    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 01-17223 Filed 7-10-01; 8:45 am]
BILLING CODE 7590-01-P