[Federal Register Volume 66, Number 201 (Wednesday, October 17, 2001)]
[Notices]
[Pages 52794-52813]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-25957]


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NUCLEAR REGULATORY COMMISSION


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

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


[[Page 52795]]



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 September 24, 2001 through October 5, 2001. 
The last biweekly notice was published on October 3, 2001 (66 FR 
50463).

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. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By November 16, 2001, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the NRC's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically from the 
Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC web site, http://www.nrc.gov/NRC/ADAMS/index.html. If a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any

[[Page 52796]]

limitations in the order granting leave to intervene, and have the 
opportunity to participate fully in the conduct of the hearing, 
including the opportunity to present evidence and cross-examine 
witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for 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. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: August 13, 2001.
    Description of amendment request: The proposed amendment would 
change the requirement to withdraw the first set of reactor vessel 
surveillance specimens by deferring withdrawal for one additional 
operating cycle.
    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:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The withdrawal in Fall 2003 refueling outage vice the March 2002 
refueling outage and the deferral of the withdrawal of the vessel 
surveillance specimens are not initiators of or precursors to any of 
the accident scenarios presented in the [Updated Safety Analysis 
Report] USAR. This schedular adjustment will not increase the 
likelihood of equipment failure, will not defeat the design reactor 
protection functions, and will not increase the likelihood of a 
catastrophic failure of any plant structure, system, or component. 
The vessel surveillance specimens are used as the basis for the 
pressure-temperature (P/T) curves. However, despite the deferral for 
one cycle of withdrawal of the vessel surveillance specimens, the P/
T curves will continue to conservatively be established in 
accordance with Regulatory Guide (RG) 1.99, ``Radiation 
Embrittlement of Reactor Vessel Materials,'' Revision 2, as 
described in the USAR. Therefore, this change does not involve an 
increase in the probability of any accident previously evaluated.
    The proposed change to the withdrawal schedule for the vessel 
surveillance specimens postpones the collection of one of two sets 
of data needed to confirm the basis of the P/T curves with no change 
to the currently allowed P/T curves. The P/T curves that are in the 
[Technical Specifications] TS will continue to be based on RG 1.99. 
The deferral of the removal of the first set of specimens will not 
affect the confirmation of the bases for the P/T curves because the 
withdrawal schedule for the second set of specimens is not being 
changed with this request. Because the basis for the P/T curves is 
maintained, this proposed change does not impact or increase the 
assumed radionuclide source term and will not result in an 
unacceptable reduction in reactor vessel toughness. Therefore, this 
change does not involve an increase in the consequences of any 
accident previously evaluated.
    In summary, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed deferral for one cycle of the removal of the vessel 
surveillance specimens does not involve a change to the plant design 
or operation. No new equipment will be installed or utilized, and no 
new operating conditions will be initiated as a result of this 
change. Because the P/T curves are not impacted, the safety function 
of the reactor vessel to mitigate the release of radioactive steam 
and limit reactor inventory loss under normal, accident, and 
transient conditions is not affected. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The deferral for one cycle of the withdrawal of the vessel 
surveillance specimens does not affect the P/T curves, and therefore 
does not affect the margin to safety for brittle fracture. Because 
two sets of specimens are needed to confirm the basis for the P/T 
temperatures and because the schedule for the withdrawal of the 
second set of specimens is not changing, the P/T curves continue in 
the interim to conform to RG 1.99. The proposed change does not 
challenge the integrity of the fuel cladding, reactor coolant 
pressure boundary that includes the reactor vessel, or the primary 
containment.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The Nuclear Regulatory Commission (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: Robert Helfrich, Mid-West Regional Operating 
Group, Exelon Generation Company, LLC, 1400 Opus Place, Suite 900, 
Downers Grove, IL 60515
    NRC Section Chief: Anthony J. Mendiola

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

    Date of amendment request: September 17, 2001.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) surveillance requirement for 
the containment spray nozzles by changing the test frequency from 
``once per 10 years'' to ``following activities that could result in 
nozzle blockage.''
    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 52797]]


    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change revises the testing requirements for the 
containment spray nozzles to only require verification that each 
spray nozzle is unobstructed following activities that could result 
in nozzle blockage. The only event for which the containment spray 
system is considered an initiator is the maximum containment 
negative pressure event. This event involves inadvertent actuation 
of containment spray following a break in the reactor water cleanup 
system inside containment described in Updated Safety Analysis 
Report (USAR) Section 6.2.1.1.4.2. This change does not increase the 
likelihood for an inadvertent actuation of the containment spray 
system.
    The proposed change does not have a detrimental impact on the 
integrity of any plant structure, system, or component that 
initiates an analyzed event. No active or passive failure mechanisms 
that could lead to an accident are affected. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident. As a result, the probability of any accident previously 
evaluated, is not significantly increased.
    The consequences of a previously evaluated accident are not 
significantly increased. The proposed change revises the current 
Surveillance Frequency from 10 years to following activities that 
could result in spray nozzle blockage. Since activities that could 
introduce foreign material into the system (such as inadvertent 
actuation of the containment spray system or loss of foreign 
material control) are the most likely cause for obstruction, testing 
or inspection following such activities would verify the nozzle(s) 
being unobstructed, and the system capable of performing its safety 
function. No other evolutions require the system boundary to be 
breached, so introduction of debris during times when maintenance 
activities are not in progress are precluded. Introduction of 
foreign materials into the system from the exterior is highly 
unlikely due to the location of the spray headers, the passive 
nature of the nozzles, and the fact that the containment spray 
headers are maintained dry which does not lend itself to active 
degradation mechanisms such as corrosion. The proposed testing 
requirements are considered sufficient to provide a high degree of 
confidence that containment spray flow will be available when 
required. Therefore, the proposed change does not significantly 
increase the consequences of an accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change to the test frequency for the containment 
spray system nozzles does not involve the use or installation of new 
equipment. Installed equipment is not operated in a new or different 
manner. No new or different system interactions are created, and no 
new processes are introduced. The current foreign material exclusion 
practices have been reviewed and judged sufficient to provide high 
confidence that debris will not be introduced during times when the 
system boundary is breached.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The revision to the containment spray nozzle testing frequency 
does not introduce any new setpoints at which protective or 
mitigative actions are initiated. No current setpoints are altered 
by this change. The design and functioning of the containment spray 
system is unchanged. Since the system is not susceptible to 
corrosion induced obstruction nor is the introduction of foreign 
material from the exterior likely, the proposed testing frequency is 
sufficient to provide high confidence that the containment spray 
system will be available to provide the flow necessary to ensure 
that the effects of drywell bypass leakage and low energy line 
breaks are mitigated. Therefore, the capacity of the system will 
remain unchanged. As a result, this 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: Robert Helfrich, Mid-West Regional Operating 
Group, Exelon Generation Company, LLC, 1400 Opus Place, Suite 900, 
Downers Grove, IL 60515
    NRC Section Chief: Anthony J. Mendiola

Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick 
Steam Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of amendment request: September 18, 2001.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) to revise the Minimum Critical 
Power Ratio (MCPR) Safety Limit values contained in TS 2.1.1.2, and 
revise the MCPR Safety Limit values from 1.10 to 1.12 for two 
recirculation loop operation and from 1.11 to 1.14 for single 
recirculation loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed license amendment will establish MCPR Safety Limit 
values of 1.12 for two recirculation loop operation and 1.14 for 
single recirculation loop operation. The revised MCPR Safety Limit 
values have been determined using NRC-approved methods and 
procedures. These procedures incorporate cycle-specific parameters 
and reduced power distribution uncertainties in the determination of 
the MCPR Safety Limit values. These proposed MCPR Safety Limit 
values do not affect the operability of any plant systems nor do 
these revised values compromise any fuel performance limits. 
Therefore, the proposed change to the MCPR Safety Limit values does 
not result in an increase in the probability of a previously 
evaluated accident.
    The consequences of a previously evaluated accident are 
dependent on the initial conditions assumed for the analysis, the 
behavior of the fuel during the accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the accident, and the setpoints at which these actions 
are initiated. The MCPR Safety Limit values are determined to ensure 
that 99.9 percent of the fuel rods will not experience boiling 
transition during any plant operation if the limit is not exceeded. 
Operational MCPR limits will be applied that ensure the MCPR Safety 
Limit is not exceeded during all modes of operation and anticipated 
operational occurrences. The MCPR Safety Limit does not impact the 
source term or pathways assumed in accidents previously evaluated. 
No analysis assumptions are violated, and there are no adverse 
effects on the factors contributing to offsite and onsite dose. The 
proposed change to the MCPR Safety Limit values does not affect the 
performance of any equipment used to mitigate the consequences of a 
previously evaluated accident. Also, the proposed change does not 
affect setpoints that initiate protective or mitigative actions. 
Based on the determination of the MCPR Safety Limit values using 
conservative NRC-approved methods and the operability of plant 
systems designed to mitigate the consequences of accidents not being 
changed, the proposed changes to the MCPR Safety Limit values does 
not significantly increase the consequences of a previously 
evaluated accident.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation. This proposed license amendment does 
not involve any facility modifications, and plant equipment will not 
be operated in a different manner. Also, no new initiating events or 
transients result from the MCPR Safety Limit changes. As a result, 
no new failure modes are being introduced. Therefore, the proposed 
changes to the MCPR Safety Limit values will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 52798]]

    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The margin of safety is established through the design of the 
plant structures, systems, and components; through the parameters 
within which the plant is operated; through the establishment of 
setpoints for actuation of equipment relied upon to respond to an 
event; and through margins contained within the safety analyses. The 
proposed change to the MCPR Safety Limit values does not adversely 
impact the performance of plant structures, systems, components, and 
setpoints relied upon to respond to mitigate an accident. The MCPR 
Safety Limit values have been calculated using NRC-approved methods 
and procedures. The MCPR Safety Limit values are determined to 
ensure that 99.9 percent of the fuel rods will not experience 
boiling transition during any plant operation if the limits are not 
exceeded, thereby ensuring that fuel cladding integrity is 
maintained. Based on the assurance that the fuel design criteria are 
being met, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New London 
County, Connecticut

    Date of amendment request: August 8, 2001.
    Description of amendment request: The proposed amendment would 
incorporate two changes into each operating license: (1) Revise the 
physical protection (security) related license condition to indicate 
that the physical security program plans listed, may, rather than do 
contain, safeguards information, and (2) change the name of the 
``Millstone Nuclear Power Station'' to the ``Millstone Power Station.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The first proposed clarification modifies the physical 
protection (security) related license condition within the 
respective operating license (OL) to indicate the physical security 
program plans listed, may, rather than do contain, safeguards 
information. The second proposed change to reflect the change in 
name of the facility from the ``Millstone Nuclear Power Station'' to 
the ``Millstone Power Station'' is editorial. Neither of these 
changes alter any regulatory requirements or have an impact on the 
acceptance criteria for any design basis accident described in the 
respective Unit 2 or 3 Updated Final Safety Analysis Report (UFSAR) 
or the Unit I Defueled Safety Analysis Report (DSAR).
    These changes have no impact on plant equipment operation. Since 
the changes are solely an administrative or editorial change to the 
OL, they cannot affect the likelihood or consequences of accidents. 
Therefore, these changes will not increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes have no impact on plant operation. Since 
the proposed changes are solely an administrative or editorial 
change to the OL, they do not affect plant operation in any way.
    The changes do not alter the plant configuration (no new or 
different type of equipment will be installed) or require any new or 
unusual operator actions. The changes do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The changes do not introduce 
any new failure modes. Therefore, the proposed changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Since the proposed changes are solely a clarification or an 
editorial change to the OL, they do not affect plant operation in 
any way. The proposed changes do not impact any acceptance criteria 
for the design basis accidents described in the respective Unit No. 
2 or No. 3 UFSAR or the Unit No. 1 DSAR and do not impact the 
consequences of accidents previously evaluated. Therefore, the 
proposed changes will not result in a reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, 
Waterford, Connecticut 06835.
    NRC Section Chief: Stephen Dembek

Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of amendment request: August 9, 2001.
    Description of amendment request: The proposed amendments modify 
the Millstone Nuclear Power Station, Unit Nos. 2 (MP2) and 3 (MP3) 
Technical Specifications (TSs) to avoid confusion between the 
qualification standards of the facility staff, who are qualified to 
American National Standards Institute (ANSI) N18.1-1971/Regulatory 
Guide (RG) 1.8 Revision 0, and the operators who will be qualified to 
the education and experience guidelines outlined by National Academy 
for Nuclear Training ACAD 00-003 ``Guidelines for Initial Training and 
Qualification of Licensed Operators.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed administrative clarification modifies the Unit Nos. 
2 and 3 TS to avoid confusion between the qualification standards of 
the facility staff, who are qualified to American National Standards 
Institute (ANSI), ``Selection and Training of Nuclear Power Plant 
Personnel,'' ANSI N18.1-1971/Regulatory Guide 1.8, Revision 0, 
``Qualification and Training of Personnel for Nuclear Power 
Plants,'' and the operators who will be qualified to the education 
and experience guidelines outlined by National Academy for Nuclear 
Training (NANT 2000 Guidelines), ACAD 00-003, ``Guidelines for 
Initial Training and Qualification of Licensed Operators.'' The 
training of the operators themselves is not affected, this change 
only modifies the education and experience requirements they must 
meet to qualify for the operator training program. The reactor 
operator and senior reactor operator applicant (or upgrade) still 
must learn and are tested on the same material, demonstrate their 
proficiency on the facility simulator and meet other requirements. 
Consequently, this change has no impact on the capability of 
licensed operators, it only modifies and provides alternative 
qualifications for entry into the program.
    This change will not alter any regulatory requirements or have 
an impact on the acceptance criteria for any design basis accident 
described in the respective Unit Nos. 2 or 3 Updated Final Safety 
Analysis Report (UFSAR).
    The change has no impact on plant equipment operation. Since the 
change is solely an administrative change to the Technical 
Specifications, it cannot affect the

[[Page 52799]]

likelihood or consequences of accidents. Therefore, this change will 
not increase the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change has no impact on plant operation. Since the 
proposed change is solely an administrative change to the Technical 
Specifications, it does not affect plant operation in any way.
    The change does not alter the plant configuration (no new or 
different type of equipment will be installed) or require any new or 
unusual operator actions. The change does not alter the way any 
structure, system, or component functions and does not alter the 
manner in which the plant is operated. The change does not introduce 
any new failure modes. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Since the proposed change is solely an administrative change to 
the Technical Specifications, it does not affect plant operation in 
any way.
    The proposed change does not impact any acceptance criteria for 
the design basis accidents described in the respective Unit Nos. 2 
or 3 UFSAR and does not impact the consequences of accidents 
previously evaluated. Therefore, the proposed change will not result 
in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    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: June 28, 2001.
    Description of amendment request: The proposed amendment modifies 
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical 
Specifications to remove the surveillance requirement associated with 
post maintenance testing of the containment isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification change to remove the 
surveillance requirement to perform post maintenance testing of the 
containment isolation valves will not cause an accident to occur and 
will not result in any change in the operation of the associated 
accident mitigation equipment. The containment isolation valves are 
not accident initiators. The proposed change will not revise the 
operability requirements (e.g., valve stroke time) for the 
containment isolation valves. Proper operation of the containment 
isolation valves will still be verified, as appropriate, following 
maintenance activities. As a result, the design basis accidents will 
remain the same postulated events described in the Millstone Unit 
No. 3 Final Safety Analysis Report, and the consequences of the 
design basis accidents will remain the same. Therefore, the proposed 
change will not increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change to the Technical Specifications does not 
impact any system or component that could cause an accident. The 
proposed change will not alter the plant configuration (no new or 
different type of equipment will be installed) or require any 
unusual operator actions. The proposed change will not alter the way 
any structure, system, or component functions, and will not 
significantly alter the manner in which the plant is operated. The 
response of the plant and the operators following an accident will 
not be different. In addition, the proposed change does not 
introduce any new failure modes. Therefore, the proposed change will 
not create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification change to remove the 
surveillance requirement to perform post maintenance testing of the 
containment isolation valves will not cause an accident to occur and 
will not result in any change in the operation of the associated 
accident mitigation equipment. The operability requirements for the 
containment isolation valves have not been changed, and proper 
operation of the containment isolation valves will still be 
verified, as appropriate, following maintenance activities. The 
containment isolation valves will continue to be able to mitigate 
the design basis accidents as assumed in the safety analysis. 
Therefore, the proposed change will not result in a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: September 7, 2001.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications Regarding the Post Accident 
Monitoring Instrumentation (Table 3.3.3-1 of Section 3.3.3. ``Post 
Accident Monitoring Instrumentation''). Specifically, the proposed 
amendment would reword the number of required channels stated for the 
core exit thermocouples (CETs) to be the same as the Standard Technical 
Specifications; delete notes that describe the redundant channels for 
the Reactor Coolant System (RCS) Hot Leg Temperature, the RCS Cold Leg 
Temperature and Main Steam Line Radiation and modify the note 
pertaining to the redundant channel for Steam Generator Level (Wide 
Range) to clarify what Condition Statements apply when the instrument 
channel and/or the Auxiliary Feedwater Flow instrument channel is 
inoperable. Other existing notes in the Table are proposed to be 
renumbered to accommodate the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Consistent with the criteria of 10 CFR 50.92, the enclosed 
[proposed] application is judged to involve no significant hazards 
based on the following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: The proposed amendment involves rewording or 
clarification of technical specification requirements to properly 
reflect the design of post accident monitoring instrumentation at 
Indian Point 3. The proposed rewording of the required channels for 
core exit thermocouples adopts the wording from the Standard 
Technical Specifications, which is applicable to the Indian Point 3 
design. The proposed deletion of Notes (a), (b), and (g) removes 
design information that is not needed for the specification to limit 
plant operation in response to inoperable instrument channels. The 
proposed rewording of Note (f) clarifies the existing requirement by 
making a more explicit statement about the applicable conditions for 
the affected functions.

[[Page 52800]]

Renumbering other Table notes is an editorial change to keep the 
notes in sequential order.
    The proposed amendment does not involve any changes to plant 
equipment, setpoints, or the way in which the plant is operated. 
These changes do not affect accident initiators or accident 
mitigating systems. Therefore, the proposed 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?
    Response: The proposed amendment involves rewording or 
clarification of technical specification requirements to properly 
reflect the design of post accident monitoring instrumentation at 
Indian Point 3. The proposed amendment does not involve any changes 
to plant equipment, setpoints, or the way in which the plant is 
operated. These changes do not affect accident initiators or 
accident mitigating systems. Therefore the proposed amendment does 
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?
    Response: The proposed amendment involves rewording or 
clarification of technical specification requirements to properly 
reflect the design and licensing basis of post accident monitoring 
instrumentation at Indian Point 3. The proposed rewording of the 
required channels for core exit thermocouples adopts the wording 
from the Standard Technical Specifications, which is applicable to 
Indian Point 3. This will ensure that appropriate condition 
statements are entered in the event that core exit thermocouples 
become inoperable. Notes (a), (b), and (g) provide design 
information that is not needed in the specification for plant 
operators to enter appropriate condition statements when inoperable 
instrument channels in the affected functions are identified. The 
rewording of Note (f) more clearly states the existing requirement 
and makes no change to the required actions or completion times for 
the associated inoperable instrument channels.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Generating Station, 600 Rocky Hill Road, Plymouth, MA 
02360.
    NRC Section Chief: Lakshminaras Raghavan, Acting.

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

    Date of amendment request: August 13, 2001.
    Description of amendment request: The proposed amendments would 
revise technical specifications (TS) to support a planned upgrade to 
the reactor water level instrumentation. Currently, many low-level 
actuation functions use Yarway level indicating switches. This includes 
emergency core cooling system (ECCS), reactor core isolation cooling 
(RCIC) and feedwater systems. The Yarways will be replaced with more 
reliable analog level transmitters and additional electronic trip 
units. The upgrade will provide sensing devices for reactor vessel 
water level signals and indications that are more reliable with less 
drift and will require less frequent surveillance requirements. The 
proposed changes align the TS surveillance requirements with the 
instrumentation upgrades. This includes changes to calibration 
frequencies, functional testing and allowable values.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    During the upcoming refueling outages at Quad Cities Nuclear 
Power Station (Unit 1 and Unit 2), a design change will be 
implemented that upgrades the existing reactor vessel level trip 
instrumentation used in various applications at Quad Cities Nuclear 
Power Station, including the Emergency Core Cooling System (ECCS), 
Reactor Core Isolation Cooling System (RCIC) and Feedwater systems.
    Technical Specification (TS) requirements that govern 
operability or routine testing of plant instruments are not assumed 
to be initiators of any analyzed event because these instruments are 
intended to prevent, detect, or mitigate accidents. Therefore, these 
changes will not involve an increase in the probability of 
occurrence of an accident previously evaluated. Additionally, these 
changes will not increase the consequences of an accident previously 
evaluated because the proposed change does not adversely impact 
structures, systems, or components (SSCs). The planned instrument 
upgrade is a more reliable design than existing equipment. The 
proposed TS change maintains existing requirements that ensure 
components are operable when necessary for the prevention or 
mitigation of accidents or transients. Revised allowable values for 
the associated functions have been established in accordance with 
EGC's setpoint methodology, which is consistent with industry 
standards. The setpoint methodology establishes TS allowable values 
that assure systems structures and components (including initiation 
and trip functions) respond in a manner consistent with the plant 
safety analysis. Furthermore, there will be no change in the types 
or significant increase in the amounts of any effluents released 
offsite. For these reasons, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes support a planned instrumentation upgrade. 
The change provides revised Surveillance Requirements to ensure 
operability. The change does not adversely impact the manner in 
which the instrument will operate under normal and abnormal 
operating conditions. These changes reflect the improved performance 
of the instrumentation upgrade and provide an equivalent level of 
safety. The changes in methods governing normal plant operation are 
consistent with the current safety analysis assumptions. Therefore, 
these changes will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change supports a planned instrumentation upgrade. 
The proposed change does not affect the probability of failure or 
availability of the affected instrumentation. The change to an 
analog trip system to monitor reactor vessel level provides for 
increased reliability. The change has no impact on the underlying 
design functions. The proposed TS surveillance requirements are 
consistent with current TS requirements for functions that employ 
analog trip unit devices. The proposed allowable values have been 
established in accordance with EGC's setpoint methodology, which 
considers instrument design and performance characteristics. The 
methodology establishes TS allowable values with sufficient margin 
to assure that the plant safety analysis assumptions (e.g., certain 
initiation and trip functions) are maintained. As such, the trip and 
actuation functions continue to ensure design basis requirements are 
maintained. Therefore, it is concluded that the proposed changes 
will not result in 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348

[[Page 52801]]

    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 29, 2001.
    Description of amendment request: The proposed license amendment 
would change the technical specifications (TSs) to reflect revised 
reactor coolant system (RCS) heatup and cooldown pressure and 
temperature (P/T) limit curves that will be valid through 22 effective 
full power years (EFPYs). The overpressure protection system (OPPS) 
power-operated relief valve (PORV) setpoints and the OPPS enabling 
temperature would also be revised. The proposed BVPS-1 P/T limits 
incorporate the results from the testing of the Capsule Y described in 
WCAP-15571, ``Analysis of Capsule Y from Beaver Valley Unit 1 Reactor 
Vessel Radiation Surveillance Program,'' Revision 0, November 2000. 
These changes have been prepared using the Nuclear Regulatory 
Commission-approved methodology described in WCAP-14040-NP-A, 
``Methodology Used to Develop Cold Overpressure Mitigating System 
Setpoints and RCS Heatup and Cooldown Limit Curves,'' Revision 2, 
January 1996, with two exceptions. These exceptions include the use of 
(1) the American Society of Mechanical Engineers (ASME) Code Case N-
640, ``Alternate Reference Fracture Toughness for Development of P-T 
Curves for Section XI, Division 1,'' March 1999, and (2) the ASME 
Boiler and Pressure Vessel Code, Section XI, ``Rule for Inservice 
Inspection of Nuclear Power Plant Components,'' Appendix G, ``Fracture 
Toughness Criteria for Protection Against Failure,'' December 1995 
(through 1996 Addendum). The TS Bases and Figure Index will also be 
changed to reflect the revisions discussed above.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes do not result in physical changes being 
made to structures, systems, or components (SSCs), or to event 
initiators or precursors. Changing the heatup and cooldown curves, 
power operated relief valve (PORV) setpoint and overpressure 
protection system (OPPS) enable temperature to reflect 22 effective 
full power years (EFPY) will not affect the ability of the OPPS to 
control the reactor coolant system (RCS) at low temperatures such 
that the integrity of the reactor coolant pressure boundary (RCPB) 
is not compromised by violating the pressure/temperature (P/T) 
limits. These changes were determined in accordance with the 
methodologies set forth in the regulations to provide an adequate 
margin of safety to ensure the reactor vessel will withstand the 
effects of normal cyclic loads due to temperature and pressure 
changes as well as the loads associated with postulated faulted 
events.
    Also, the proposed changes do not impact the design of plant 
systems such that previously analyzed SSCs would now be more likely 
to fail. The initiating conditions and assumptions for accidents 
described in the Updated Final Safety Analysis Report (UFSAR) remain 
as previously analyzed. Thus, the proposed changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed changes do not alter any assumptions previously 
made in the radiological consequence evaluations nor affect 
mitigation of the radiological consequences of an accident described 
in the UFSAR. As such, the consequences of accidents previously 
evaluated in the UFSAR will not be increased and no additional 
radiological source terms are generated. Therefore, there will be no 
reduction in the capability of those SSCs in limiting the 
radiological consequences of previously evaluated accidents and 
reasonable assurance that there is no undue risk to the health and 
safety of the public will continue to be provided. Thus, the 
proposed changes do not involve a significant increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed changes do not involve physical changes to 
analyzed SSCs or changes to the modes of plant operation defined in 
the technical specification. The proposed changes do not involve the 
addition or modification of plant equipment (no new or different 
type of equipment will be installed) nor do they alter the design or 
operation of any plant systems. No new accident scenarios, accident 
or transient initiators or precursors, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes.
    The proposed changes do not cause the malfunction of safety-
related equipment assumed to be operable in accident analyses. No 
new or different mode of failure has been created and no new or 
different equipment performance requirements are imposed for 
accident mitigation. As such, the proposed changes have no effect on 
previously evaluated accidents.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed changes have been determined through supporting 
analyses to be in accordance with the methodologies set forth in the 
regulations. Compliance with NRC approved methodologies provide for 
an adequate margin of safety and ensure the reactor vessel will 
withstand the effects of normal cyclic loads due to temperature and 
pressure changes as well as the loads associated with postulated 
faulted events as described in the UFSAR.
    The new heatup and cooldown curves define the limits for 
ensuring prevention of nonductile failure for the BVPS Unit No. 1 
reactor vessel and do not significantly reduce the margin of safety 
for the plant.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: L. Raghavan (Acting).

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

    Date of amendment request: September 27, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to (1) change the diesel fuel 
supply volume required for diesel generator (DG) operability, (2) 
clarify existing wording, (3) add a TS limiting condition for operation 
(LCO) and a TS surveillance requirement (SR) regarding DG air 
receivers, (4) delete a current TS SR concerning DG starting air 
compressors, and (5) restructure and renumber the TS LCOs and SRs for 
applicability and administrative purposes.
    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 a[n] accident 
previously evaluated.
    The proposed Technical Specification changes do not introduce 
new equipment or new equipment operating modes, nor do the proposed 
changes alter existing system

[[Page 52802]]

relationships. The proposed amendment does not introduce new failure 
modes.
    The proposed revision to the Monticello TS[s] renumbers and 
relocates TS[s] as appropriate to provide a more understandable TS, 
deletes an existing TS SR which does not satisfy the requirements of 
10 CFR 50.36 for inclusion in the TS[s], adds a new TS LCO and SR 
for DG air start receivers which more appropriately complies with 
the requirements of 10 CFR 50.36, and revises the minimum number of 
gallons of diesel fuel required in the Diesel Oil Storage Tank for 
the DG to be declared operable.
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 introduce a new mode of plant 
operation, or involve a physical modification to the plant. The 
proposed Technical Specification changes do not introduce new 
equipment, nor do the proposed changes alter existing system 
relationships. The proposed amendment does not introduce new failure 
modes.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes maintain the current TS requirements for 
safe operation of the Monticello plant. The proposed changes do not 
involve a physical modification to the plant, or a new mode of 
operation. The proposed changes do not alter the scope of equipment 
currently required to be operable nor do the proposed changes affect 
equipment safety functions. The proposed Technical Specification 
changes do not introduce new equipment, nor do the proposed changes 
alter existing system relationships. The proposed amendment does not 
introduce new failure modes.
    Therefore, these proposed changes will 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: William D. Reckley

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 amendment request: August 17, 2001.
    Description of amendment request: The proposed Technical 
Specifications (TS) change will: (1) modify Salem TS surveillance 
requirement 4.6.2.3, and (2) revise the associated TS Bases. 
Specifically, the proposed change will modify the current acceptance 
criterion for the service water flow rate through the Containment Fan 
Coil Units from  2,550 gallons per minute (gpm) to 
 2,300 gpm.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The containment ventilation system, including the containment fan 
coil units is not an accident initiator.
    The proposed TS change to modify the Salem TS surveillance 
requirement 4.6.2.3 to the service water-cooling water flow through the 
fan coil units is bounded by the present licensing and design bases 
analyses. The new proposed flow rate, in conjunction with its 
associated heat exchanger thermal fouling factor, will continue to 
maintain the assumed minimum containment heat removal capability to be 
within the Salem Updated Final Safety Analysis Report (UFSAR) Chapter 
15 analyses. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed TS change to modify the Salem TS surveillance 
requirement 4.6.2.3 to the service water-cooling water flow though the 
fan coil units is bounded by the present licensing and design bases 
analyses. The manner and frequency at which the surveillance test is 
conducted remains unchanged. The physical facility remains unchanged.
    Therefore, the new proposed flow rate does not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. Does not involve a significant reduction in a margin of safety.
    The proposed TS change to modify the Salem TS surveillance 
requirement 4.6.2.3 to the service water-cooling water flow though the 
fan coil units is bounded by the present licensing and design bases 
analyses. The new proposed flow rate, in conjunction with its 
associated heat exchanger thermal fouling factor, will continue to 
maintain the assumed minimum containment heat removal capability to be 
within the Salem UFSAR Chapter 15 analyses. Consequently, the existing 
margins of safety with respect to the current design-basis assumptions 
of pressure of 47 psig and a saturation temperature of 271  deg.F in 
containment during a design-basis accident is maintained.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear 
Plant, Unit 1, Appling County, Georgia

    Date of amendment request: August 31, 2001.
    Description of amendment request: The proposed amendment would 
allow a one-time deferral of the Type A Containment Integrated Leak 
Rate Test based on the risk-informed guidance in Regulatory Guide 
1.174.
    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 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 fifteen (15) 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 which the

[[Page 52803]]

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. HNP Unit 1 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 the containment 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. 
Therefore, the proposed Technical Specification change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed 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 TS 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.
    HNP Unit 1 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 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: August 31, 2001.
    Description of amendment request: The proposed amendments would 
extend the completion times for the required actions associated with 
restoring an inoperable emergency diesel generator.
    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 changes extend the Technical Specifications 
required Completion Times for restoration of an inoperable emergency 
diesel generator (DG) to a maximum of 14 days. Additionally, the 
proposed extension of the Completion Time to 14 days results in a 
corresponding extension of the time period associated with discovery 
of failure to meet Limiting Condition for Operation 3.8.1 to 17 
days. (This provides a maximum time limit for overlapping 
inoperabilities of DGs and offsite sources.)
    For both Plant Hatch units A and C DGs, to utilize the 72 hours 
to 14 day period of the proposed extended Completion Time, 
compensatory action is required to ensure two DGs per unit remain 
available. This action consists of dedicating the 1B DG to that unit 
with the inoperable DG. This means that the 1B DG will be inhibited 
from an automatic swap to the opposite unit when that unit (the non-
maintenance unit) experiences and undervoltage condition on its F 
4160 volt bus, regardless of the presence or absence of a loss of 
coolant accident (LOCA) signal. Inhibiting the automatic transfer 
makes the 1B DG inoperable (with a Completion Time of 14 days) for 
the non-maintenance unit.
    Completion Times are not an initial condition or assumption of 
any analyzed event. DGs are not initiators of any analyzed event. No 
event mitigation assumes more than two DGs per unit. The 
consequences of an accident are independent of the time the DGs are 
out of service provided adequate DG availability is assured. 
Compensatory actions are proposed in this amendment request that 
ensure adequate DG availability for both Plant Hatch units. 
Therefore, the assumptions regarding DG available are maintained.
    To fully evaluate the effect of the proposed DG Completion Time 
extension, Probabilistic Safety Assessment methods and a 
deterministic analysis were utilized. The

[[Page 52804]]

results of the analyses show no significant increase in Core Damage 
Frequency (CDF) and Large Early Release Frequency (LERF).
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an event previously 
analyzed.
    2. Do the proposed changes create the possibility of a new or 
different type of accident from any previously evaluated?
    The proposed changes do involve a change to the plant 
configuration when either unit's A or C DG is utilizing the extended 
Completion Time (i.e., inoperable in excess of 72 hours). That 
configuration change ensures that both units have two dedicated DGs. 
Furthermore, affixing the 1B DG to one unit will cause it (1B DG) to 
be inoperable with respect to the Technical Specifications. Ensuring 
two DGs available to each unit for event mitigation in no way 
creates the possibility of a new or different type of accident.
    No other change in the design, configuration, or method of 
operation of the plant is introduced by the proposed change. The 
changes do not alter any assumptions made in the safety analyses. No 
new failure modes are introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Since all assumptions of the plant event analyses are 
maintained, there is no effect on the margin of safety in any safety 
analyses. If there is any margin of safety ascribed to DG 
availability and plant risk, it has been determined that such a 
margin of safety is not significantly reduced, as the proposed 
changes have been evaluated both deterministically and using a risk-
informed approach. These evaluations concluded the following with 
respect to the proposed changes:
    Applicable regulatory requirements will continue to be met, 
adequate defense-in-depth will be maintained, sufficient safety 
margins will be maintained, and any increases in CDF and LERF are 
small and consistent with the NRC Safety Goal Policy Statement 
(Federal Register, Vol. 51, p. 30028 (51 FR 30028), August 4, 1986, 
as interpreted by NRC Regulatory Guides 1.174 and 1.177). 
Furthermore, increases in risk posed by potential combinations of 
equipment out of service during the proposed DG extended Completion 
Time will be managed by the site configuration risk management 
procedure, consistent with 10 CFR 50.65, ``Requirements for 
Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants,'' paragraph (a)(4).
    The availability of offsite power together with the availability 
of the other DGs and the use of on-line risk assessment tools 
provide adequate compensation for the potential small incremental 
increase in plant risk of the extended DG Completion Time. In 
addition, the increased availability of the DGs during refueling 
outages offsets the small increase in plant risk during operation. 
The proposed extended DG Completion Times, in conjunction with the 
availability of the other DGs continues to provide adequate 
assurance of the capability to provide power to the engineered 
safety features buses. Therefore, implementation of the proposed 
changes will 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: September 7, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 3.6.11, ``Ice Bed,'' 
Surveillance Requirement (SR) 3.6.11.2, SR 3.6.11.3, and the associated 
Bases, to lower the minimum required average ice basket weight from 
1236 pounds to 1110 pounds, and the corresponding total weight of the 
stored ice in the ice condenser from 2,403,800 pounds to 2,158,000 
pounds.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The primary purpose of the ice bed is to provide a large heat 
sink to limit peak containment pressure in the event of a release of 
energy from a design basis loss-of-coolant (LOCA) or high energy 
line break (HELB) in containment. The LOCA requires the greatest 
amount of ice compared to other accident scenarios, therefore the 
reduction in ice weight is based on the LOCA analysis. The amount of 
ice in the bed has no impact on the initiation of an accident, but 
rather on the mitigation of the accident.
    The containment integrity analysis shows that the proposed 
reduced ice weight is sufficient to maintain the peak containment 
pressure below the containment design pressure, and that the 
containment heat removal systems function to rapidly reduce the 
containment pressure and temperature in the event of a LOCA. 
Therefore, containment integrity is maintained and the consequences 
of an accident previously evaluated in the Updated Final Safety 
Analysis Report (UFSAR) are not significantly increased.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The ice condenser serves to limit the peak pressure inside 
containment following a LOCA. TVA has evaluated the revised 
containment pressure analysis and determined that sufficient ice 
would be present to maintain the peak containment pressure below the 
containment design pressure. Therefore, the reduced ice weight does 
not create the possibility of an accident that is different than any 
already evaluated in the WBN UFSAR. No new accident scenarios, 
failure mechanisms, or limiting single failures are introduced as a 
result of this proposed change.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The containment integrity analysis for reduced ice weight 
results in a peak containment pressure that is slightly lower than 
that in the previous analysis of record. This reduction in peak 
pressure, along with the ice weight reduction, is due to the removal 
of analytical conservatism combined with a better segmental 
representation of the mass and energy release transient from the 
computer models.
    The revised technical specifications ice weight surveillance 
limits are based on the ice weight assumed in the containment 
integrity analysis, with margin included for sublimation that is 
based on actual sublimation data from the first three refueling 
cycles at WBN. The analysis further demonstrates that the existing 
relationship between ice bed melt-out and containment spray 
switchover has been conservatively maintained. With the reduced ice 
inventory, melt-out of the ice bed following a worst case large 
break LOCA has been determined to occur after the switchover of 
containment spray to the recirculation mode. Thus, the reduced ice 
bed mass does not result in a reduction in the margin for operator 
action to effect the switchover.

    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.

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

    Date of amendment request: August 24, 2001.
    Brief description of amendments: The proposed change would revise 
Comanche Peak Steam Electric Station,

[[Page 52805]]

Units 1 and 2, Technical Specification (TS) 3.3.2, entitled ``ESFAS 
[Engineered Safety Features Actuation System] Instrumentation,'' and TS 
3.3.6, entitled ``Containment Ventilation Isolation Instrumentation,'' 
to change the surveillance frequency for Westinghouse Electric Company-
type AR relays, used as Solid State Protection System slave relays or 
auxiliary relays, from quarterly to refueling outage frequency. 
Surveillance Requirements 3.3.2.6 and 3.3.6.5 would be revised to 
change the frequency from ``92 days'' to ``92 days OR 18 months for 
Westinghouse type AR relays.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the Technical Specifications does not 
result in a condition where the design, material, or 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 or function 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 the 
performance of any accident mitigation systems or alter any 
assumptions previously made in evaluating the radiological 
consequences of an accident as described in the safety analysis 
report.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not alter the performance of the ESFAS 
mitigation systems assumed in the plant safety analysis. Changing 
the interval for periodically verifying ESFAS slave relays (assuring 
equipment operability) will not create any new accident initiators 
or scenarios.
    Therefore, the proposed change does not create the possibility 
of 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?
    Response: No.
    The proposed 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.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.

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

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

    Date of amendment request: September 27, 2001 (WO 01-0038)
    Description of amendment request: The proposed amendment would 
revise Section 5.3.1.1 of the Technical Specifications to replace the 
current qualifications in ANSI/ANS 3.1-1981 for licensed operators and 
senior operators with the National Academy for Nuclear Training, 
``Guidelines for Initial Training and Qualification of Licensed 
Operators,'' dated January 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 proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS change is an administrative change to clarify 
the current requirements for licensed operator qualifications and 
licensed operator training program. [The change conforms] to the 
current requirements of 10 CFR 55.
    Although licensed operator qualifications and training may have 
an indirect impact on accidents previously evaluated, the NRC 
considered this impact during the rulemaking process, and by 
promulgation of the revised 10 CFR 55 rule, concluded that this 
impact remains acceptable as long as the licensed operator training 
program is certified to be accredited and is based on a systems 
approach to training. WCNOC's [Wolf Creek Nuclear Operating 
Corporation's] licensed operator training program is accredited by 
INPO [Institute for Nuclear Power Operations] and is based on a 
system[']s approach to training. The proposed TS change takes credit 
for the INPO accreditation of the licensed operator training 
program. The TS requirements for all other unit staff qualifications 
remain unchanged.
    Therefore, the proposed change does not involve a signification 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS change is an administrative change to clarify 
the current requirements for licensed operator qualifications and 
licensed operator training program and to conform to the revised 10 
CFR 55.
    As noted above, although licensed operator qualifications and 
training may have an indirect impact on the possibility of a new or 
different kind of accident from any accident previously evaluated, 
the NRC considered this impact during the rulemaking process, and by 
promulgation of the revised [10 CFR 55] rule, concluded that this 
impact remains acceptable as long as the licensed operator training 
program is certified to be accredited and based on a system[']s 
approach to training. As previously noted, WCNOC's licensed operator 
training program is accredited by INPO and is based on a system[']s 
approach to training. The proposed TS change takes credit for the 
INPO accreditation of the licensed operator training program. The TS 
requirements for all other unit staff qualifications remain 
unchanged.
    Additionally, the proposed TS change does not affect plant 
design, hardware, system operation, or procedures. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed TS change is an administrative change to clarify 
the current requirements applicable to licensed operator 
qualifications and licensed operator training program. This change 
is consistent with the requirements of 10 CFR 55. The TS 
qualification requirements for all other unit staff remain 
unchanged.
    Licensed operator qualifications and training can have an 
indirect impact on a margin of safety. However, the NRC considered 
this impact during the rulemaking process, and by promulgation of 
the revised 10 CFR 55 [rule], determined that this impact remains 
acceptable when licensees maintain a licensed operator training 
program that is accredited and based on a system[']s approach to 
training. As noted previously, WCNOC's licensed operator training 
program is accredited by INPO and is based on a system[']s approach 
to training.
    The NRC has concluded, as stated in NUREG-1262, ``Answers to 
Questions at Public Meetings Regarding Implementation of Title 10, 
Code of Federal Regulations, Part 55 on Operators' Licenses,'' that 
the standards and guidelines applied by INPO in their training 
accreditation program are equivalent to those put forth or endorsed 
by the NRC. As a result, maintaining an INPO accredited, systems 
approach based licensed operator training program is equivalent to 
maintaining an NRC approved licensed operator training program which 
conform with applicable NRC Regulatory Guides or

[[Page 52806]]

NRC endorsed industry standards. The margin of safety is maintained 
by virtue of maintaining an INPO accredited licensed operator 
training program.
    In addition, the NRC has recently published NRC Regulatory Issue 
Summary 2001-01, ``Eligibility of Operator License Applicants,'' 
dated January 18, 2001, ``to familiarize addresses with the NRC's 
current guidelines for the qualification and training of reactor 
operator (RO) and senior operator (SO) license applicants.'' This 
document again acknowledges that the INPO National Academy for 
Nuclear Training (NANT) guidelines for education and experience, 
outline acceptable methods for implementing the NRC's regulations in 
this area.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

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

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

    Date of amendment request: August 14, 2001, as supplemented on 
August 21, 2001.
    Description of amendment request: The proposed amendment would 
extend the allowed outage time for the high pressure coolant injection 
(HPCI) and reactor core isolation cooling systems from 7 days to 14 
days. Requirements were added to immediately assure the availability of 
alternate means of high pressure coolant makeup. Also clarifying 
changes were made to Technical Specification (TS) 3.5.E.2 and TS 
3.5.G.2 by reformatting the TSs to make nomenclature consistent 
regarding HPCI and the automatic depressurization system (ADS) as being 
systems not subsystems.
    Date of publication of individual notice in Federal Register: 
September 16, 2001 (66 FR 48152).
    Expiration date of individual notice: October 18, 2001.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: January 29, 2001, as 
supplemented July 6, 2001.
    Brief description of amendment: The amendment removes the note from 
TMI-1 Technical Specification 4.5.4.1 which restricts the applicability 
of the specified Engineered Safeguards Feature (ESF) Systems leakage 
rate limit of 15 gallons per hour to the current operating cycle (Cycle 
13). The amendment also approves full scope implementation of an 
alternate source term for TMI-1 in accordance with Title 10 of the Code 
of Federal Regulations (10 CFR) Section 50.67.
    Date of issuance: September 19, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 235.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31703).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 19, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 23, 2001, as 
supplemented August 22 and September 17, 2001.
    Brief description of amendment: The amendment revised the TMI-1 
Technical Specification requirements for containment integrity 
associated with the personnel and emergency air locks during fuel 
movement and refueling operations. Partial implementation of an 
alternate source term (AST) in accordance with Regulatory Guide 1.183, 
``Alternate Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Plants,'' and Title 10 of the Code of Federal 
Regulations, Section 50.67,

[[Page 52807]]

which the licensee had also requested in its application, was not 
necessary because the Commission approved full implementation of an AST 
for TMI-1 in Amendment No. 235 dated September 19, 2001.
    Date of issuance: October 2, 2001.
    Effective date: As of its date of issuance, contingent upon the 
licensee's implementation of regulatory commitments contained in the 
licensee's letters dated August 22 and September 17, 2001, and shall be 
implemented within 30 days of issuance.
    Amendment No.: 236.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31702).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 2, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 6, 2000, as 
supplemented July 13 and September 6, 2001.
    Brief description of amendment: The amendment revises the once-
through steam generator (OTSG) surveillance criteria contained in the 
TMI-1 Technical Specifications (TSs) to allow OTSG tubes to remain in 
service with indications of inside diameter intergranular attack 
located below the upper tubesheet secondary face. The changes also 
extend the repair criteria from a cycle-to-cycle basis to a permanent 
basis.
    Date of issuance: October 5, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 237.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7669).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 15, 2001.
    Brief description of amendments: The amendments delete Technical 
Specifications Section 5.5.3, ``Post Accident Sampling System,'' for 
Palo Verde Nuclear Generating Station, Units Nos. 1, 2 and 3, and 
thereby eliminate the requirements to have and maintain the post-
accident sampling system.
    Date of issuance: September 28, 2001.
    Effective date: September 28, 2001, and shall be implemented within 
7 months of the date of issuance.
    Amendment Nos.: Unit 1-136, Unit 2-136, Unit 3-136.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41611).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 28, 2001.
    No significant hazards consideration comments received: No.

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: May 1, 2001, as supplemented 
August 20, 2001.
    Brief description of amendments: The amendments change the 
Technical Specifications related to the pressure-temperature limit 
curves.
    Date of issuance: October 4, 2001.
    Effective date: October 4, 2001.
    Amendment Nos.: 214 and 241.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29350). The August 20, 2001, supplement contained clarifying 
information only, and did not change the initial no significant hazards 
consideration determination, or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 4, 2001.
    No significant hazards consideration comments received: No.

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: January 17, 2001, as 
supplemented March 23 and August 31, 2001.
    Brief description of amendments: The amendments change the 
Technical Specifications to relax the 24-month surveillance frequency 
of excess flow check valves (EFCVs) by limiting the number of tests to 
a representative sample every 24 months such that each EFCV will be 
tested at least once every 10 years.
    Date of issuance: October 4, 2001.
    Effective date: October 4, 2001.
    Amendment Nos.: 215 and 242.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11052). The March 23 and August 31, 2001, supplements contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 4, 2001.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-400, Shearon Harris 
Nuclear Power Plant (HNP), Wake and Chatham Counties, North Carolina

    Date of application for amendment: December 13, 2000, as 
supplemented on February 9, and August 3, 2001.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3/4.9.2, ``Refueling Operations--Instrumentation,'' 
and the associated Bases to permit using one Source Range Nuclear Flux 
Monitor and one Wide Range Neutron Flux Monitor during MODE 6 
(Refueling) instead of the two Source Range Nuclear Flux Monitors 
specified in the current HNP TS.
    Date of issuance: September 10, 2001.
    Effective date: September 10, 2001.
    Amendment No. 105.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7672).
    The February 9, and August 3, 2001, submittals contained clarifying 
information only, and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 10, 2001.
    No significant hazards consideration comments received: No.

[[Page 52808]]

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: December 14, 2000, as 
supplemented August 16, and September 12, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification 3/4.8.1 related to Emergency Diesel Generators (EDGs), 
and specifically revises Surveillance Requirement 4.8.1.1.2.f.7, the 
24-hour EDG endurance run test, by removing the restriction to perform 
the test during shutdown conditions.
    Date of issuance: October 3, 2001.
    Effective date: October 3, 2001.
    Amendment No.: 106.
    Facility Operating License No. NPF-63: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7673). The August 16, and September 12, 2001 supplements contained 
clarifying information that did not change the scope of the December 
14, 2000, application nor the proposed initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 3, 2001.
    No significant hazards consideration comments received: No.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix, County, Michigan

    Date of amendment request: October 26, 2000, as supplemented by 
letters dated February 9, February 28, March 14, March 15, March 23, 
May 2, July 13, July 17, and August 2, 2001.
    Brief description of amendment: The amendment revises the Defueled 
Technical Specifications to reflect the removal of the original 75-ton 
Reactor Building gantry crane and its replacement with an upgraded 
single-failure proof crane.
    Date of issuance: September 28, 2001.
    Effective date: The license amendment is effective as of its date 
of issuance and shall be implemented within 60 days from the date of 
issuance.
    Amendment No.: 122.
    Facility Operating License No. DPR-6: The amendment revised the 
Defueled Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22025). The supplemental letters dated May 2, July 13, July 17, and 
August 2, 2001, provided additional clarifying information, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 28, 2001.
    No significant hazards considerations comments received: No.

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

    Date of application for amendment: December 29, 2000, as 
supplemented May 2 and July 19, 2001.
    Brief description of amendment: The amendment revises the Fermi 2 
Technical Specifications associated with handling irradiated fuel 
assemblies, based on reevaluation of the design-basis fuel handling 
accident analysis with an alternative radiological source term.
    Date of issuance: September 28, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 144.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9381). The application was renoticed on August 27, 2001 (66 FR 45062), 
due to supplemental information beyond the scope of the initial notice. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 28, 2001.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of application for amendment: March 9, 2001, as supplemented 
by letters dated July 25, September 10, and September 13, 2001.
    Brief description of amendment: The amendment revised the cold leg 
elbow tap flow coefficients used in the determination of Reactor 
Coolant System flow rate. There are no changes to the associated 
Technical Specifications with this amendment.
    Date of issuance: October 2, 2001.
    Effective date: As of its date of issuance and shall be implemented 
before the startup of Cycle 12, and will be in effect only for the 
duration of Cycle 12.
    Amendment No.: 186.
    Facility Operating License No. NPF-52: Amendment did not revise the 
Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34281). The supplements dated July 25, September 10, and September 13, 
2001, provided clarifying information that did not change the scope of 
the March 9, 2001, application nor the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 2, 2001.
    No significant hazards consideration comments received: No.

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 application of amendments: July 18, 2000, supplemented 
August 22 and November 8, 2000, and June 7, July 26, and September 5, 
2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications to incorporate provisions of the Automatic 
Feedwater Isolation System.
    Date of Issuance: September 26, 2001.
    Effective date: As of the date of issuance. It shall be implemented 
for each Oconee unit prior to reactor startup following installation of 
the system and training of appropriate personnel.
    Amendment Nos.: 320, 320, and 320.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56949). The supplements dated August 22 and November 8, 2000, and 
June 7, July 26, and September 5, 2001, provided clarifying information 
that did not change the scope of the July 18, 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 September 26, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 24, 2001, as supplemented by 
letter dated September 24, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to incorporate the provisions to perform routine 
diesel generator (DG) monthly testing by gradually accelerating the DG 
to

[[Page 52809]]

operating speed. In addition, a new TS was added to require fast starts 
of the DGs on a 184-day frequency.
    Date of issuance: September 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
120 days from the date of issuance.
    Amendment No.: 121.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13801). The supplemental letter dated September 27, 2001, provided 
additional information that did not expand the scope of the NRC staff's 
initial proposed no significant hazards consideration determination (66 
FR 13801, published March 7, 2001).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 23, 2001, as supplemented by letters 
dated July 23, 2001, and August 23, 2001.
    Brief description of amendment: The amendment changes the following 
Technical Specifications (TSs): (1) the value of the safety limit 
minimum critical power ratio was changed in TS 2.1.1.2, (2) an 
editorial clarification to TS 5.6.5.a.5) was added to include the 
applicable reactor protection system instrumentation function, and (3) 
the list of the approved methodologies in TS 5.6.5.b. and the 
associated Bases and References were updated.
    Date of issuance: October 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 122.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34281). The supplemental letters dated July 23, 2001, and August 23, 
2001, provided additional information that did not expand the scope of 
the application or change the staff's initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 3, 2001.
    No significant hazards consideration comments received: No.

Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: May 31, 2001.
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.6, ``Technical Specification (TS) Bases Control 
Program,'' to provide consistency with the changes to 10 CFR 50.59 
which were published in the Federal Register (64 FR 53582) on October 
4, 1999.
    Date of issuance: October 2, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 192.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38761). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 7, 2000, as 
supplemented December 29, 2000.
    Brief description of amendment: The changes revise Technical 
Specification Section 3.7.B.4 to allow a one-time replacement of 
Station 125V DC batteries 31 and 32 while at power. The one-time change 
is necessary to support an on-line replacement of the existing 
batteries with new batteries. In addition, a change is made on a one-
time basis to conduct testing the battery while the plant is not 
shutdown. Also included is an administrative change involving the 
deletion of an expired one-time limiting condition for operation 
statement related to an Emergency Diesel Generator Fuel Oil Storage 
Tank repair effort.
    Date of issuance: September 19, 2001.
    Effective date: September 19, 2001.
    Amendment No.: 208.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (66 
FR 15922).
    The December 29, 2000, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 19, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 31, 2001.
    Brief description of amendment: The amendment revised and 
transferred the inservice testing portion of Technical Specification 
(TS) 4.0.5 to TS 6.5.8, and eliminated the inservice inspection portion 
of TS 4.0.5. In addition, other sections of the TSs that reference TS 
4.0.5 were revised to be consistent with the revisions discussed above.
    Date of issuance: September 24, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 233.
    Facility Operating License No. NPF-6: Amendment revised the TSs.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44167).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 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: May 22, 2001.
    Brief description of amendment: The amendment changes Technical 
Specifications (TS) 3/4.7.1.2, Emergency Feedwater System, and expands 
and clarifies the current TS.
    Date of issuance: October 4, 2001.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 173.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications and Surveillance Requirements.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34283).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 2001.
    No significant hazards consideration comments received: No.

[[Page 52810]]

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

    Date of application for amendments: April 23, 2001.
    Brief description of amendments: The amendments deleted the loose 
parts monitoring system from the LGS Units 1 and 2 Technical 
Specifications and Bases. The amendments were based on the conclusions 
of the Boiling Water Reactor Owners' Group TopicalReport NEDC-32975P, 
``Regulatory Relaxation for BWR Loose Parts Monitoring System,'' which 
was approved by the Nuclear Regulatory Commission's Safety Evaluation 
dated January 25, 2001.
    Date of issuance: As of date of issuance and shall be implemented 
within 30 days.
    Effective date: September 19, 2001.
    Amendment Nos.: 153 and 117.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41619).
    The Commission's related evaluation of the amendments is contained 
in a safety evaluation dated September 19, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 30, 2001 (two letters), as 
supplemented July 24 (two letters), and August 13, 2001.
    Brief description of amendment: This amendment revises Technical 
Specification 5.5.12 to allow a one-time change in the containment 
integrated leak rate test interval from the current 10 years to a test 
interval of 15 years.
    Date of issuance: October 4, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 244.
    Facility Operating License No. DPR-56: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36341). The July 24 (two letters), and August 13, 2001, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the application beyond the scope of the original Federal Register 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 28, 2001, as supplemented 
by letters dated May 18, June 15, and July 18, 2001.
    Brief description of amendment: The amendment approves changes to 
the BVPS-1 Technical Specification boron concentration limits for the 
refueling water storage tank, accumulators, boron injection tank (BIT), 
and the reactor coolant system/refueling canal during Mode 6. In 
conjunction with the reduction in the maximum boron concentration in 
the BIT, the temperature controls on the BIT are eliminated.
    Date of issuance: September 24, 2001.
    Effective date: Immediately and to be implemented within 60 days.
    Amendment No: 242.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38763). The May 18, June 15, and July 18, 2001, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the initial Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 28, 2001.
    Brief description of amendment: These amendments approved 
reductions in the reactor coolant system and secondary coolant system 
specific activity limits specified in TS 3/4.4.8, ``Reactor Coolant 
System Specific Activity,'' and TS 3/4.7.1.4, ``Plant Systems 
Activity.'' These TS changes support revised safety analyses of the 
design-basis main steam line break dose consequence analysis, which 
assumes higher primary-to-secondary accident induced leakage in 
accordance with the methodology described in Generic Letter 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
by Outside Diameter Stress Corrosion Cracking.'' These amendments also 
authorized Updated Final Safety Analysis Report (UFSAR) changes.
    Date of issuance: September 28, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 244.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications and authorized changes to the UFSAR.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29354).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 28, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 18, 2001, as 
supplemented by letters dated February 20, April 12, May 7, May 18, 
June 9 (3 letters), June 26, June 29, August 21, and September 5, 2001.
    Brief description of amendment: A portion of this amendment 
approves revisions to BVPS-2 TS 3/4.4.9, ``Pressure/Temperature 
Limits,'' heatup and cooldown curves.
    Date of issuance: September 24, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 122.
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 27, 2001 (66 FR 
39211). The February 20, April 12, May 7, May 18, June 9 (3 letters), 
June 26, June 29, August 21, and September 5, 2001, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
beyond the scope of the original notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 24, 2001.
    No significant hazards consideration comments received: No.

[[Page 52811]]

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 application for amendments: July 18, 2001, as supplemented 
August 30 and September 6, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.9.4 and its associated Bases to allow the 
containment equipment door to be open during core alterations or 
movement of non-recently irradiated fuel within the containment, 
provided that the capability for closure is maintained.
    Date of issuance: September 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 216 and 210.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TS.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41622). The August 30 and September 6, 2001, submittals provided 
clarifying information that did not change the scope of the July 18, 
2001, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 26, 2000, as 
supplemented February 1, 2001, June 29, 2001, and August 10, 2001.
    Brief description of amendments: The amendments would approve 
changes to revise the current licensing basis, as stated in the updated 
final safety analysis report, to require operator action to mitigate 
the effects of a loss of seal injection cooling to the reactor coolant 
pumps.
    Date of issuance: September 28, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 255 and 238.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
approve changes to the updated final safety analysis report.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62386) The February 1, June 29, and August 10, 2001, supplemental 
letters did not change the scope of the proposed action and did not 
change the NRC's preliminary no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 28, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 30, 2001, as supplemented 
June 27 and August 3, 2001.
    Brief description of amendment: The amendment conforms the license 
to reflect the transfer of Facility Operating License No. DPR-43 for 
the Kewaunee Nuclear Power Plant (KNPP) to the extent held by Madison 
Gas & Electric Company (MG&E) to Wisconsin Public Service Corporation 
(WPSC), as approved by Order of the Commission dated September 20, 
2001.
    Date of issuance: September 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 159.
    Facility Operating License No. DPR-43: Amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 27, 2001 (66 FR 
39214). The August 3, 2001, supplement was within the scope of the 
initial application as originally noticed. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
September 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 8, 2000.
    Brief description of amendments: The amendments revised the 
requirements for the containment isolation valves in the hydrogen/
oxygen analyzer containment penetrations. The related safety evaluation 
also provided approval of an associated request to use closed system 
boundary valves that do not completely meet the guidance described in 
the Standard Review Plan, Section 6.2.4, ``Containment Isolation 
System.''
    Date of issuance: September 28, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 195 & 170.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31713). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 28, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 12, 2000, as 
supplemented April 9, 2001.
    Brief description of amendment: This amendment changes the 
Technical Specifications (TSs) associated with the drywell vacuum 
breakers and the suppression pool vacuum breakers to provide 
consistency between the Hope Creek TSs and the improved standard TSs 
(NUREG-1433).
    Date of issuance: October 3, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 133.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71137). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 3, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 17, 2001, as supplemented on 
August 6, August 17, and September 12, 2001
    Brief description of amendment: The amendment revises the Technical 
Specifications to permit an increase in the allowable leak rate for the 
Main Steam Isolation Valves (MSIVs) and to delete the MSIV Sealing 
System. These changes are based on the use of an alternate source term 
and the guidance provided in Regulatory Guide 1.183, ``Alternate 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors.''
    Date of issuance: October 3, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented during Refueling Outage 10, currently scheduled to commence 
in October 2001.
    Amendment No.: 134.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.

[[Page 52812]]

    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34288). The letters dated August 6, August 17, and September 12, 2001, 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 3, 2001.
    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: September 26, 2000, as 
supplemented on October 6, 2000, and May 21, 2001.
    Brief description of amendments: The amendments modify the Salem 
Technical Specifications by increasing the as-found setpoint tolerance 
for the Pressurizer Safety Valves from 1% to 
3%; increasing the as-found setpoint tolerance for the Main 
Steam Safety Valves (MSSV) from 1% to 3%; 
changing the required actions for inoperable MSSVs; and removing 
specifications and references related to plant operation with three 
Reactor Coolant System loops.
    Date of issuance: September 19, 2001
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 244 and 225
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (65 
FR 69065), as superseded on August 8, 2001 (66 FR 41624).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 19, 2001.
    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: May 31, 2000, as supplemented 
on August 2, 2001.
    Brief description of amendments: The amendments modify the Salem 
Technical Specifications (TSs) Surveillance Requirements for: (1) The 
Control Room Envelope Air Conditioning System (CREACS), (2) the 
Auxiliary Building Ventilation System (ABVS), and (3) the Fuel Handling 
Building Ventilation System (FHVS). Salem TSs will now require the use 
of American Society for Testing and Materials (ASTM) D3803-1989, 
``Standard Test Method for Nuclear-Grade Activated Carbon,'' as the 
test protocol to evaluate charcoal samples from the ABVS, CREACS, and 
FHVS.
    Date of issuance: September 19, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 245 and 226.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46014). The August 2, 2001, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 19, 2001.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: June 29, 2001 as supplemented 
by letter dated August 20, 2001.
    Brief description of amendments: Revise Technical Specifications 
(TSs) 3.7.10, ``Emergency Chilled Water (ECW)'' and 3.7.11, ``Control 
Room Emergency Air Cleanup System (CREACUS)'' and the associated TSs 
Bases. The proposed change would revise the Allowed Outage Time for a 
single inoperable train of both the ECW and CREACUS from 7 days to 14 
days.
    Date of Issuance: October 4, 2001.
    Effective date: October 4, 2001, to be implemented within 30 days 
of issuance
    Amendment Nos.: Unit 2-181; Unit 3-172
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44175). The August 20, 2001 supplemental letter provided additional 
clarifying information, did not expand the scope of the proposed 
amendment as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 4, 2001.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: June 29, 2000, as supplemented August 
31, 2001.
    Brief description of amendments: The amendments revise design bases 
in the Final Safety Analysis Report. The change adds a description of 
the methodology Southern Nuclear Operating Company uses to determine 
what systems and components need to be protected from tornado missiles.
    Date of issuance: September 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 150 and 142.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48758).
    The supplement dated August 31, 2001, provided clarifying 
information that did not change the scope of the June 29, 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 September 26, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 4, 2000, as supplemented by 
letters dated April 30, June 18, and July 18, 2001.
    Brief description of amendments: The amendments revise the 
Technical Specifications to increase the spent fuel storage capacity 
from 2,026 to 3,373 fuel assemblies in the spent fuel pool.
    Date of issuance: October 2, 2001.
    Effective date: As of the date of issuance and shall be implemented 
no later than January 31, 2002.
    Amendment Nos.: 87/87.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 4, 2000 (65 FR 
75737).
    The April 30, June 18, and July 18, 2001, supplemental letters 
provided clarifying information that was within the scope of the 
original Federal Register notice and did not change the staff's initial 
no significant hazards consideration determination.

[[Page 52813]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 2, 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: April 17, 2001.
    Brief description of amendment: The amendment removes unnecessary 
details for certain secondary post-accident monitoring instrumentation 
from Technical Specification Table 3.2.6.
    Date of Issuance: October 2, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 204.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 16, 2001 (66 FR 
27178).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 2, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 23, 2001 (CO 01-0013).
    Brief description of amendment: The amendment deletes (1) certain 
license conditions from Facility Operating License No. NPF-42, and (2) 
reporting requirements in Table 5.5.9-2, ``Steam Generator Tube 
Inspection,'' in Section 5.5.9, ``Steam Generator (SG) Tube 
Surveillance Program,'' of the technical specifications. License 
Conditions 2.C.(4), and 2.C.(6) through 2.C.(14), Section 2.F, and 
Attachments 2 and 3 to Facility Operating License No. NPF-42 are 
deleted, and the list of the attachments and appendices to Facility 
Operating License No. NPF-42 is revised to reflect the deletion of the 
attachments. The reporting requirements deleted in Table 5.5.9-2 
duplicate requirements in 10 CFR 50.72.
    Date of Issuance: September 24, 2001.
    Effective date: September 24, 2001, and shall be implemented within 
60 days from the date of issuance.
    Amendment No.: 141.
    Facility Operating License No. NPF-42: The amendment revised the 
operating license and the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22035).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 2001
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 15, 2000, and supplements 
dated October 3, 2000, and September 13, 2001.
    Brief description of amendment: The amendment revises footnotes (b) 
and (c) of Table 1.1-1, ``Modes,'' and adds a program plan to Section 
5.5, ``Programs and Manuals,'' of the Wolf Creek Generating Station 
Technical Specifications. The amendment will allow the plant to operate 
at full power with one closure bolt less than fully tensioned for one 
operating cycle.
    Date of issuance: September 27, 2001.
    Effective date: September 27, 2001, to be implemented within 60 
days from the date of issuance.
    Amendment No.: 142.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59227).
    The supplements dated October 3, 2000, and September 13, 2001, 
provided additional clarifying information, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 2001.
    No significant hazards consideration comments received: No.


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



    Dated at Rockville, Maryland, this 10th of October 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-25957 Filed 10-16-01; 8:45 am]
BILLING CODE 7590-01-P