[Federal Register Volume 67, Number 14 (Tuesday, January 22, 2002)]
[Notices]
[Pages 2917-2935]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-1210]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 28, 2001 through January 10, 2002. 
The last biweekly notice was published on January 8, 2002 (67 FR 924).

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

[[Page 2918]]

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's Public Document 
Room (PDR), 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 February 21, 2002, 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 PDR, 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/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, 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 PDR, 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/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

[[Page 2919]]

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

    Date of amendments request: November 9, 2001.
    Description of amendments request: The amendments would revise 
Technical Specification (TS) 5.6.5b to add topical report CENPD-404-P-
A, ``Implementation of ZIRLOTM Cladding Material in CE 
Nuclear Power Fuel Assembly Designs,'' to the list of analytical 
methods used to determine core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change allows the use of methods required for the 
implementation of ZIRLOTM clad fuel rods in PVNGS [Palo 
Verde Nuclear Generating Station], Units 1, 2, and 3. The use of this 
methodology will not increase the probability of an accident because 
the plant systems will not be operated outside of design limits, no 
different equipment will be operated, and system interfaces will not 
change.
    As ZIRLOTM material is introduced to the reactor, 
transition cores will exist in which ZIRLO'' and Zircaloy-4 clad fuel 
assemblies are co-resident. Fuel assemblies clad with each material 
will be evaluated based on the approved topical reports.
    The use of this additional methodology will not increase the 
consequences of an accident because Limiting Conditions of Operation 
(LCOs) will continue to restrict operation to within the regions that 
provide acceptable results, and Reactor Protection System (RPS) trip 
setpoints will restrict plant transients so that the consequences of 
accidents will be acceptable. In addition, the consequences of the 
accidents will be calculated using NRC accepted methodologies.
    The transition cores that will exist as ZIRLO'' clad fuel is 
introduced to the reactor will not increase the consequences of an 
accident. Operation within the LCOs and RPS setpoints will continue to 
restrict plant transients so that the consequences of accidents will be 
acceptable.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with any existing equipment, alter the equipment's function, 
or change the method of operating the equipment. The proposed change 
does not alter plant conditions in a manner that could affect other 
plant components. The proposed change does not cause any existing 
equipment to become an accident initiator. The ZIRLOTM clad 
fuel rod design does not introduce features that could initiate an 
accident.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    Safety Limits ensure that Specified Acceptable Fuel Design Limits 
(SAFDLs) are not exceeded during steady state operation, normal 
operational transients and anticipated operational occurrences. All 
fuel limits and design criteria shall be met based on the approved 
methodologies defined in the topical reports. The RPS in combination 
with the LCOs will continue to prevent any anticipated combination of 
transient conditions for reactor coolant system temperature, pressure, 
and thermal power level that would result in a violation of the Safety 
Limits. Therefore, the proposed changes will have no impact on the 
margins as defined in the Technical Specification bases.
    The safety analyses determine the LCO settings and RPS setpoints 
that establish the initial conditions and trip setpoints, which ensure 
that the Design Basis Events (Postulated Accidents and Anticipated 
Operational Occurrences) analyzed in the Updated Final Safety Analysis 
Report (UFSAR) produce acceptable results. In addition, all fuel limits 
and design criteria shall be satisfied. The Design Basis Events that 
are impacted by the implementation of ZIRLO'' cladding will be analyzed 
using the NRC accepted methodology described in CENPD-404-P-A.
    The change in the fuel rod cladding material and the use of the 
ECCS [emergency core cooling system] performance evaluation models, 
CENPD-132, Supplement 4-P, ``Calculative Methods for the CE Nuclear 
Power Large Break LOCA Evaluation Model'' and CENPD-137, Supplement 2-
P, ``Calculative Methods for the CE Small Break LOCA Evaluation Model'' 
will not involve a reduction in the margin of safety because LCOs and 
Limiting Safety System Settings (LSSS) will be adjusted, if necessary, 
to maintain acceptable results for the impacted Design Basis Events.
    Therefore, 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 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendments request: December 13, 2001.
    Description of amendments request: The amendments would add (1) the 
phrase ``or if open, capable of being closed'' to Limiting Condition 
for Operation 3.9.3, ``Containment Penetration,'' and (2) Surveillance 
Requirement (SR) 3.9.3.3 on verifying the capability to close the 
equipment hatch, if open. For refueling operations, the amendments 
would allow the equipment hatch to be open during core alterations and 
movement of irradiated fuel assemblies inside containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment[s] to Technical Specification (TS) 3.9.3 
``Containment Penetrations,'' would allow the equipment hatch to 
remain open, but capable of being closed, during CORE ALTERATIONS or 
movement of irradiated fuel assemblies within containment. The 
position of the equipment hatch (open or closed) is not an initiator 
of any accident.
    The fuel handing accident (FHA) contained in the Updated Final 
Safety Analysis Report, Revision 11, currently assumes that the 
entire

[[Page 2920]]

airborne radioactivity reaching the containment is released to the 
outside environment. This results in a maximum offsite dose of 74.7 
rem to the thyroid and 0.39 rem to the whole body. The calculated 
control room dose of 11.5 rem thyroid and 0.13 whole body are within 
the acceptance criteria specified in General Design Criteria 19 
``Control Room.''
    Therefore, the proposed amendment request does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to TS 3.9.3 ``Containment Penetrations,'' 
allowing the equipment hatch to be open and capable of being closed 
does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. Thus, the proposed 
amendment request does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed amendment to TS 3.9.3 ``Containment Penetrations,'' 
allowing the equipment hatch to be open and capable of being closed 
remains bounded by previously determined radiological dose 
consequences for a FHA inside containment. The previously analyzed 
dose consequences were determined to be within the limits of 10 CFR 
[part] 100, ``Reactor Site Criteria,'' and they meet the acceptance 
criteria of NUREG-0800 Standard Review Plan for the Review of Safety 
Analysis Reports for Nuclear Power Plants Section 15.7.4 
``Radiological Consequences of Fuel Handling Accidents.'' Therefore, 
the proposed amendment request does not involve a significant 
reduction in a margin of safety. Additionally, a new surveillance 
will be added to verify the capability to close the equipment hatch, 
if open and CORE ALTERATIONS or movement of irradiated fuel 
assemblies are in progress within containment, at a frequency of 
seven days.
    Based on the above, APS [the licensee] concludes that the 
activities associated with the proposed amendment(s) present no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c) and, accordingly, a finding of ``no significant 
hazards consideration'' is justified.

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

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: October 1, 2001.
    Description of amendment request: The proposed amendment would 
modify the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical 
Specifications (TSs) to increase the emergency diesel generator (EDG) 
allowed outage time (AOT), to perform a verification of the offsite 
circuits within 1 hour prior to or after entering the condition of 
either an inoperable offsite source or inoperable EDG, to revise the 
requirements for the pressurizer heaters and the pressurizer power 
operated relief and block valves, and to improve the format of the 
electrical power sources action requirements. The Bases of the affected 
TSs will be modified to address the proposed 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. The NRC staff reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's analysis, which is 
based on the representation made by the licensee in the October 1, 
2001, application, is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS changes to increase the EDG AOT, to perform a 
verification of the offsite circuits within 1 hour prior to or after 
entering the condition of either an inoperable offsite source or 
inoperable EDG, to revise the requirements for the pressurizer 
heaters and the pressurizer power operated relief and block valves, 
and to improve the format of the electrical power sources action 
requirements are not accident initiators nor will they impact the 
consequences of any previously evaluated accidents. Therefore, the 
proposed changes will not increase the probability or consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the TSs do not impact any system or 
component that could cause an accident nor do the proposed changes 
alter the plant configuration or require any unusual operator 
actions or alter the way any structure, system, or component 
functions. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The TS changes to revise the requirements for the pressurizer 
heaters and the pressurizer power operated relief and block valves, 
the TS change to allow verification of offsite circuits within 1 
hour prior to or after entering the condition of an inoperable 
offsite source or inoperable EDG, and the changes to improve the 
format of the electrical power sources action requirements do not 
change the TS-required safety limits or safety system settings; 
therefore, these additional changes will not result in a significant 
reduction in a margin of safety. The proposed TS changes to increase 
the EDG AOT do not affect any assumptions or inputs to the safety 
analyses. Unavailability of a single EDG due to maintenance or 
repair activities does not reduce the number of EDGs below the 
minimum required to mitigate all DBAs. Therefore, the proposed 
change will not result in a significant reduction in a margin of 
safety.
    Based on this analysis, it appears that the three standards of 
10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.

    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: August 6, 2001.
    Description of amendment request: The proposed amendments would 
decrease the McGuire Units 1 & 2 and Catawba Unit 1 Overtemperature 
Delta Temperature (OTT) Allowable Values and the McGuire Units 
1 & 2 and Catawba Units 1 & 2 Overpower Delta Temperature 
(OPT) Allowable Values. OTT and OPT are trip 
functions provided in the reactor trip system to protect against 
departure from nucleate boiling and to ensure fuel integrity under all 
overpower conditions. The licensee states that due to changes in reload 
reactor core designs since the 1.0 deg.F hot leg streaming uncertainty 
value was determined, it is now necessary to increase the uncertainty 
value to 1.21 deg.F. Associated changes in the TS Table 3.3.1-1 
OTT and OPT allowable values have been proposed. The 
licensee states that the decreases are in the conservative direction 
and will not adversely affect the steady-state or transient analyses 
documented in the Updated Final Safety Analysis Reports. In addition, 
the licensee has proposed

[[Page 2921]]

two minor editorial changes for Catawba Units 1 & 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

First Standard

    Would implementation of the changes proposed in this LAR involve 
a significant increase in the probability or consequences of an 
accident previously evaluated?
    No. This license amendment request [LAR] proposes to decrease 
the McGuire Units 1 & 2 and Catawba Unit 1 Overtemperature Delta 
Temperature (OTT) Allowable Values and the McGuire Units 1 
& 2 and Catawba Units 1 and 2 Overpower Delta Temperature 
(OPT) Allowable Values. This decrease is in the 
conservative direction and will not adversely affect the steady-
state or transient analyses documented in the Updated Final Safety 
Analysis Report. These changes have no impact on accident 
probabilities or consequences.
    The proposed changes to the Catawba Table of Contents and Bases 
are solely administrative in nature and have no impact on any 
accidents.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.

Second Standard

    Would implementation of the changes proposed in this LAR create 
the possibility of a new or different kind of accident from any 
previously evaluated?
    No. The proposed changes contained in this LAR only correct 
administrative errors and add conservative operability requirements 
which are consistent with the plants' existing licensing bases. No 
new or different kinds of accidents are being created.
    3. Involve a significant reduction in margin of safety.

Third Standard

    Would implementation of the changes proposed in this LAR involve 
a significant reduction in a margin of safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. These barriers are unaffected by the changes proposed in 
this LAR. Consequently, no margin of safety will be significantly 
impacted by this LAR. Conclusion
    Based upon the preceding evaluation, performed pursuant to 
10CFR50.92, Duke Energy Corporation has concluded that 
implementation of this LAR at McGuire and Catawba Nuclear Station 
will not involve a significant hazards consideration. The changes 
proposed in this LAR make a conservative decrease in the McGuire 
Units 1 & 2 and Catawba Unit 1OTT Allowable Values and the 
McGuire Units 1 & 2 and Catawba Units 1 and 2 OPT Allowable 
Values and correct unrelated administrative errors. Following 
implementation of these proposed changes, McGuire and Catawba will 
continue to be operated in a conservative manner.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: December 20, 2001.
    Description of amendment request: The proposed amendments would 
change the McGuire Technical Specifications (TS) to eliminate the 
revision number and dates from the list of topical reports that contain 
the analytical methods used to determine the core operating limits. 
This proposed change is consistent with the NRC approved Industry 
Technical Specifications Task Force (TSTF) Standard Technical 
Specifications Traveler TSTF-363, ``Revise Topical Report References in 
ITS 5.6.5 COLR''. Implementation of the changes proposed in this 
license amendment request will have no adverse impact on Duke's 
practices for controlling the methodologies used to develop the core 
operating limits for McGuire. The complete citations (i.e., report 
number, title, revision number, report date or NRC safety evaluation 
date, and any supplements) for each of the topical reports listed in TS 
5.6.5 will be displayed as applicable in each station's Core Operating 
Limits Report (COLR). NRC review and approval of new or revised topical 
reports will continue to be obtained in the same manner. Changes to the 
COLRs will be controlled by 10 CFR 50.59.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


    Duke Energy Corporation has made the determination that this 
license amendment request (LAR) for McGuire Technical Specifications 
(TS) involves No Significant Hazards. This determination was made 
through the application of the standards established by 10CFR50.92. 
The three standards are discussed below.
    1. Would implementation of the changes proposed in this LAR 
involve a significant increase in the probability or consequences of 
an accident previously evaluated?
    No. This LAR makes an administrative change to TS 5.6.5.b, Core 
Operating Limits Report (COLR), affecting a list of documents that 
are separately reviewed and approved by the NRC. The changes 
proposed to TS 5.6.5.b have no substantive impact on the McGuire 
licensing bases. Only NRC-approved methodologies will be used to 
generate the core operating limits. Based on these considerations, 
it has been determined that the proposed changes have no impact on 
any accident probabilities or consequences.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. This LAR makes administrative changes that have no impact on 
any accident analyses.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The analytical methodologies used to generate the core 
operating limits are unchanged by this LAR. As such, this LAR has no 
affect on margins of safety. Future changes to these methodologies 
will remain subject to NRC review and approval. Therefore, this 
proposed amendment does not involve a reduction in any margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: December 20, 2001.
    Description of amendment request: The proposed amendments would 
make changes in the Technical Specifications (TS) Bases Control Program 
to reflect changes in the NRC's regulations in 10 CFR 50.59 as noticed 
in the Federal Register on October 4, 1999. The proposed changes in the 
license amendment request are consistent with

[[Page 2922]]

an NRC approved Technical Specifications Task Force Standard TS 
Traveler (TSTF-364).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
[license amendment request] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR makes an administrative change to the Technical 
Specifications [TS] made necessary as part of Duke's implementation 
of revised NRC regulations. The changes proposed to these TS have no 
substantive impact on the McGuire licensing bases, nor Duke's 
ability to conservatively evaluate changes to these licensing bases. 
Therefore, the proposed changes have no impact on any accident 
probabilities or consequences.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. This LAR makes administrative changes that have no impact on 
any accident analyses.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The proposed changes are administrative, an implementation 
of the revised 10CFR50.59 regulation. Implementation of the revised 
10CFR50.59 regulation provides the necessary regulatory requirements 
to ensure that nuclear plants' margin of safety is preserved.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: October 16, 2001.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to incorporate changes 
resulting from the use of an alternate source term and the 
implementation of several plant modifications. The proposed TS changes 
include the following:
     The Penetration Room Ventilation System (PRVS) will be 
removed from TS because the PRVS will not be credited in licensing 
analyses that determine Control Room and off-site doses.
     The Spent Fuel Pool Ventilation System (SFPVS) will be 
removed from TS because the SFPVS will not be credited in licensing 
analyses that determine Control Room and off-site doses.
     During certain refueling operations, the containment air 
locks and/or the equipment hatch and penetrations providing direct 
access from the containment atmosphere to the outside atmosphere will 
be permitted to be unisolated under administrative controls. 
Additionally, the requirement to maintain an operable automatic 
isolation capability for the Reactor Building Purge system during 
refueling will be removed from TS.
     The allowable value for the Reactor Building leakage rate 
will be lowered from 0.25 w%/day to 0.20 w%/day.
     The requirement to measure Reactor Building leakage in 
excess of 50% of La to the penetration room will be removed 
from TS.
     The Ventilation Filter Testing Program (VFTP) will be 
revised to remove all references to the PRVS and SFPVS and their 
testing requirements.
     The VFTP acceptance criterion for the Control Room 
Ventilation System Booster Fan trains will be revised to require 
 97.5% radioactive methyl iodide removal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The AST [Alternative Source Term] and those plant systems 
affected by implementing the proposed changes to the TS are not 
assumed to initiate design basis accidents. The AST does not affect 
the design or operations of the facility. Rather, the AST is used to 
evaluate the consequences of a postulated accident. The 
implementation of the AST has been evaluated in the revisions to the 
analysis of the design basis accidents for Oconee Nuclear Station. 
Based on the results of these analyses, it has been demonstrated 
that, with the requested changes, the dose consequences of these 
events meet the acceptance criteria of 10 CFR 50.67 and Regulatory 
Guide 1.183. 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 
evaluated.
    The AST and those plant systems affected by implementing the 
proposed changes to the TS are not assumed to initiate design basis 
accidents. The systems affected by the changes are used to mitigate 
the consequences of an accident that has already occurred. The 
proposed TS changes and modifications do not significantly affect 
the mitigative function of these systems. Consequently, these 
systems do not alter the nature of events postulated in the Safety 
Analysis Report nor do they introduce any unique precursor 
mechanisms. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The implementation of the AST, proposed changes to the TS and 
the implementation of the proposed modifications have been evaluated 
in the revisions to the analysis of the consequences of the design 
basis accidents for the Oconee Nuclear Station. Based on the results 
of these analyses, it has been demonstrated that with the requested 
changes the dose consequences of these events meet the acceptance 
criteria of 10 CFR 50.67 and Regulatory Guide 1.183. Thus, the 
proposed amendment 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: December 20, 2001.
    Description of amendment request: The proposed amendments would 
revise the licensing basis associated with the failure of non-Category 
I (non-seismic) piping.
    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.

[[Page 2923]]

    No. The License Amendment Request (LAR) proposes to change the 
licensing basis for non-Category I (non-seismic) piping to assume a 
through-wall crack as the postulated piping failure. The proposed 
change does not involve any physical alteration of plant systems, 
structures or components, changes in parameters governing normal 
plant operation, or methods of operation. The proposed change does 
not affect any Chapter 15 accident analyses. Duke evaluated the 
effects of flooding caused by a leak from a crack size calculated 
using the SRP [Standard Review Plan] guidelines in the 16-inch HPSW 
[High Pressure Service Water] header. This evaluation concluded that 
for the bounding case, the effects of flooding can be mitigated 
without adversely affecting safety-related equipment. At least an 
hour is available from detection for operator action to isolate the 
leak. Therefore, the probability or consequences of an accident 
previously evaluated is not significantly increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No. The License Amendment Request (LAR) changes the licensing 
basis associated with non-seismic moderate energy line breaks. The 
proposed change does not necessitate a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in parameters governing normal plant operation. Therefore, 
the possibility of a new or different kind of accident from any kind 
of accident previously evaluated is not created.
    3. Involve a significant reduction in a margin of safety.
    No. The License Amendment Request (LAR) changes the licensing 
basis associated with non-seismic moderate energy line breaks. The 
impact of flooding from a seismically induced crack in non-seismic 
moderate energy piping has been evaluated. Adequate time exists for 
operator action to isolate flooding sources prior to adversely 
affecting safety-related equipment required for safe shutdown. As 
such, the proposed change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: December 20, 2001.
    Description of amendment request: The proposed amendments would 
make changes in the Technical Specifications (TS) to eliminate the use 
of the term ``unreviewed safety question.'' The change is proposed by 
the licensee to reflect changes in the NRC's regulations in 10 CFR 
50.59 as noticed in the Federal Register on October 4, 1999. The 
proposed changes in the license amendment request are consistent with 
an NRC approved Technical Specifications Task Force Standard TS 
Traveler (TSTF-364).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
[license amendment request] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR makes an administrative change to the Technical 
Specifications [TS] made necessary as part of Duke's implementation 
of revised NRC regulations. The changes proposed to these TS have no 
substantive impact on the Oconee licensing bases, nor Duke's ability 
to conservatively evaluate changes to these licensing bases. 
Therefore, the proposed changes have no impact on any accident 
probabilities or consequences.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. This LAR makes administrative changes that have no impact on 
any accident analyses.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The proposed changes are administrative, an implementation 
of the revised 10 CFR 50.59 regulation. Implementation of the 
revised 10 CFR 50.59 regulation provides the necessary regulatory 
requirements to ensure that nuclear plants' margin of safety is 
preserved.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard J. Laufer, Acting.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick (JAF) Nuclear Power Plant, Oswego County, New York

    Date of amendment request: November 2, 2001.
    Description of amendment request: The proposed change to the JAF 
Nuclear Power Plant Technical Specifications establishes a combined 
leakage rate limit for the sum of the four main steam line leakage 
rates that is equal to four times the current main steam line valve 
(MSIV) leakage rate limit.
    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:

    Operation of the JAF plant in accordance with the proposed 
amendment would not involve a significant hazards consideration as 
defined in 10 CFR [Code of Federal Regulations] 50.92 since it would 
not: Involve an increase in the probability or consequences of an 
accident previously evaluated.
    The proposed amendment does not involve a change to structures, 
components, or systems that would affect the probability of an 
accident previously evaluated in the FitzPatrick Updated Final 
Safety Analysis Report (UFSAR).
    The proposed amendment results in no change in radiological 
consequences of the design basis LOCA [loss-of-coolant accident] as 
currently analyzed for the FitzPatrick Plant. These analyses were 
calculated assuming a combined total MSIV leakage at accident 
pressure for determining acceptance to the regulatory limits for the 
offsite, control room, and Technical Support Center (TSC) radiation 
doses as contained in 10 CFR 100 and 10 CFR 50, Appendix A, GDC 19 
[General Design Criteria]. The proposed change does not compromise 
existing radiological equipment qualification, since the combined 
total MSIV leakage rate has been factored into existing equipment 
qualification analyses for 10 CFR 50.49.
    Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not modify the MSIVs or any other plant 
system or structure associated with this amendment and therefore, 
will not affect their capability to perform their design function. 
The combined total main steam line leakage rate is included in the 
current radiological analyses for the assessment of radiation 
exposure following an accident.
    This proposal changes the allowable leakage rate from a per 
valve limit to a total combined leakage rate limit for all four main 
steam lines but does not change the cumulative limit. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    Involve a significant reduction in a margin of safety.

[[Page 2924]]

    The leakage rate limit specified for the MSIVs is used to 
quantify the maximum amount of bypass leakage assumed in the LOCA 
radiological analysis. Results of the analysis are evaluated against 
the dose guidelines contained in GDC 19 and 10 CFR 100. The margin 
of safety in this context is considered to be the difference between 
the calculated dose exposures and the guidelines provided by the GDC 
19 and 10 CFR 100. Therefore, since the proposed combined total main 
steam line leakage rate limit is unchanged from the assumed maximum 
leakage rate for MSIVs, for the purpose of calculating potential 
radiation dose, the margin of safety is not affected because the 
postulated radiation doses remain the same.

    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. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: December 21, 2001.
    Description of amendment request: Exelon proposed changes that 
would revise Technical Specification (TS) 2.1 to incorporate revised 
Safety Limit Minimum Critical Power Ratios (SLMCPRs) due to the cycle-
specific analysis performed by Global Nuclear Fuel for LGS Unit 1, 
Cycle 10, which will include the use of the GE-14 fuel product line.
    Basis for proposed no significant hazards consideration 
determination: As required by Section 50.91(a) of Title 10 of the Code 
of Federal Regulations (CFR), 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 or 10 CFR 
50.92(c). The NRC staff's review is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The derivation of the cycle-specific SLMCPRs for incorporation 
into the TSs, and its use to determine cycle-specific thermal 
limits, has been performed using the methodology discussed in 
``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-14-US, 
June 2000, which incorporates Amendment No. 25. Amendment No. 25 
provides the methodology for determining the cycle-specific MCPR 
safety limits that replaces the former generic fuel type dependent 
values. Amendment No. 25 was approved by the NRC (Nuclear Regulatory 
Commission) in a March 11, 1999, safety evaluation.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling. The probability of fuel damage will 
not increase as a result of this change. Likewise, the consequences 
of accidents previously evaluated are not affected by the revised 
SLMCPRs values. Therefore, the proposed TS change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, calculated to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core if the limit is not violated. SLMCPRs are based on a 
calculation using an NRC-approved methodology discussed in NEDE-
24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-14-US, 
June 2000. The SLMCPR is not an accident initiator, and its revision 
will 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 new SLMCPRs are calculated using NRC-approved methodology 
discussed in NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-14-US, June 2000. This methodology uses the same standards 
and margins that were used in the former generic fuel type methodology. 
Therefore, the proposed TS change will not involve a significant 
reduction in a margin of safety previously approved by the NRC.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Clifford.

National Aeronautics and Space Administration (NASA), Docket No. 50-30, 
Plum Brook Reactor Facility (PBRF), Sandusky, Ohio

    Date of amendment request: December 20, 1999, as supplemented on 
March 26, November 19, and December 20, 2001.
    Description of amendment request: The proposed amendment would 
allow decommissioning of the Plum Brook Test Reactor Facility.
    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 approval of the PBRF Decommissioning Plan 
involve a significant increase in the probability or consequences of 
an accident previously evaluated?
    All nuclear fuel has been removed from the PBRF site. 
Radioactive inventories at the PBRF are very small compared to those 
in operating reactors (both power and non-power) and in various 
kinds of fuel cycle facilities subject to NRC regulation. Analyses 
indicate that decommissioning activities would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated in the current Final Hazards Summary 
for the NASA Plum Brook Reactor Facility.
    Summary: NASA considers that the approval of the Decommissioning 
Plan does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed approval of the PBRF Decommissioning Plan 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    The current Final Hazards Summary for the NASA Plum Brook 
Reactor Facility evaluated those cause-and-effect accidents related 
to external events and loss/failure of reactor support systems that 
would result in the dispersal of fission products and radioactive 
materials to the environment. Due to the combined absence of fuel at 
the PBRF site and the non-operational condition of reactor support 
systems, NASA has determined that decommissioning activities, as 
described in the Decommissioning Plan, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Summary: NASA considers that the approval of the Decommissioning 
Plan does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Current Technical Specifications adequately restrain the scope 
and nature of decommissioning activities to loose equipment removal 
and preparations for dismantlement. Approval of the proposed 
Decommissioning Plan provides for additional controls prior to 
commencement of dismantlement activities, thereby achieving a 
greater margin of safety.
    Summary: NASA considers that the approval of the Decommissioning 
Plan does not involve a significant reduction in a margin of safety.
    Based on the above evaluations, NASA concludes that the 
activities associated with

[[Page 2925]]

the above described changes present no significant hazards 
consideration under the standards set forth in 10 CFR 50.92(c) and, 
accordingly, a finding by the NRC of no significant hazards 
consideration is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for the Licensee: J. William Sikora, Esquire, 21000 
Brookpark Road, Mail Stop 500-118, Cleveland, OH 44135.
    NRC Section Chief: Patrick M. Madden.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station Unit No. 2, Oswego County, New York

    Date of amendment request: October 5, 2001; as revised on January 
4, 2002. This notice supersedes a previous notice (66 FR 55020) 
published on October 31, 2001, which was based upon the licensee's 
application dated October 5, 2001.
    Description of amendment request: The licensee proposed to amend 
the Technical Specifications (TSs) to change the licensing basis 
requirement for establishing containment hydrogen monitoring ``within 
30 minutes'' to ``within 3 hours'' of initiating emergency core cooling 
following a loss-of-coolant accident (LOCA). The January 4, 2002, 
revision reduces the proposed delay from 3 hours to 90 minutes.
    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 operation of Nine Mile Point Unit 2 in accordance with 
the proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The Updated Safety Analysis Report (USAR) Chapter 15 accident 
analyses do not require or take credit for hydrogen monitoring to be 
established shortly after a loss of coolant accident (LOCA). Post-
LOCA hydrogen production occurs over a long period of time, and an 
extension from ``30 minutes'' to ``90 minutes'' for establishing 
hydrogen monitoring will have a positive impact on the ability of 
the operators to concentrate on their more immediate actions while 
having no negative impact on containment integrity or the long-term 
assessment efforts. Therefore, the proposed license amendment will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) The operation of Nine Mile Point Unit 2 in accordance with 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Control room operators use the containment hydrogen monitors 
following a LOCA to establish hydrogen control measures should it 
become necessary. The proposed license amendment would not eliminate 
the requirement to establish hydrogen monitoring, but would allow it 
to be delayed until those actions required to mitigate the accident 
and verify proper operation of essential safety equipment have been 
completed. The proposed extension maintains the requirement to 
establish hydrogen monitoring well before calculated conditions 
inside the containment indicate any need to initiate hydrogen 
control measures. Therefore, the proposed license amendment will not 
create a new or different kind of accident from any accident 
previously evaluated.
    (3) The operation of Nine Mile Point Unit 2 in accordance with 
the proposed amendment will not involve a significant reduction in a 
margin of safety.
    The need to establish hydrogen control measures will not be 
present within the first 90 minutes following a LOCA since there 
will not be significant hydrogen accumulation. By extending the time 
allowed to establish containment hydrogen monitoring, the operators 
can remain focused on the actions necessary to mitigate the accident 
before directing their attention to hydrogen control measures and 
other long-term actions. The proposed extension maintains the 
requirement to establish hydrogen monitoring well before calculated 
conditions inside the containment indicate any need to initiate 
hydrogen control measures. Therefore, the proposed license amendment 
will 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington DC 20005-3502.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: December 21, 2001.
    Description of amendment request: The proposed changes would revise 
Technical Specification (TS) Surveillance Requirements (SR) 4.3.1.2 and 
4.3.2.2 to allow verification in place of demonstration of response 
time associated with certain pressure sensors, differential pressure 
sensors, process protection racks, nuclear instrumentation, and logic 
systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to TS 3/4.3.1 and TS 3/4.3.2 do not result 
in a condition where the design, material, and construction 
standards that were applicable prior to the proposed changes are 
altered. The same Reactor Trip System (RTS) and Engineered Safety 
Features Actuation System (ESFAS) instrumentation is being used; the 
response time allocations/modeling assumptions in the Seabrook 
Station UFSAR [updated final safety analysis report] analyses are 
still the same; only the method of verifying time response is 
changed. The proposed change will not modify any system interface 
and will not increase the probability or consequences of an accident 
previously evaluated since these events are independent of this 
change.
    The proposed changes do not affect the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated in the 
Seabrook Station UFSAR. Further, the proposed changes do not 
increase the types and amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures.
    Therefore, it is concluded that these proposed revisions to TS 
3/4.3.1 and TS 3/4.3.2 do not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes to TS 3/4.3.1 and TS 3/4.3.2 do not alter 
the performance of the pressure and differential pressure sensors 
used in the plant protection systems, nor do the proposed changes 
alter the performance of the Process Protection racks, Nuclear 
Instrumentation, and Logic Systems used in the plant protection 
systems. The sensors will still have their response time verified by 
test before placing the sensor in operational service and after any 
maintenance that could affect response time; and the plant 
protection systems will still have response time verified by test 
before being placed in operational service.
    For the pressure and differential pressure sensors; and for the 
Process Protection racks, the Nuclear Instrumentation, and the Logic 
Systems; changing the method of periodically verifying instrument 
response from time

[[Page 2926]]

response testing to calibration and channel checks (assuring 
equipment operability) will not create any new accident initiators 
or scenarios.
    The periodic calibration of the pressure and differential 
pressure sensors will detect significant degradation in the sensor 
response characteristic.
    The periodic calibration of the Process Protection racks, the 
Nuclear Instrumentation, and the Logic Systems will continue to be 
used to detect significant degradation that could cause the response 
time characteristic to exceed the total allowance. The total time 
response allowance for each function bounds degradation that cannot 
be detected by the periodic surveillance.
    Thus, these proposed revisions to TS 3/4.3.1 and TS 3/4.3.2 do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes to TS 3/4.3.1 and TS 3/4.3.2 do not affect 
the total system response time assumed in the Seabrook Station UFSAR 
analyses. The periodic system response time verification method for 
the pressure and differential pressure transmitters; and the 
periodic system response time verification method for the Process 
Protection racks, the Nuclear Instrumentation, and the Logic 
Systems, is modified to allow use of actual test data or engineering 
data. The method of verification will continue to provide assurance 
that the total system response is within that defined in Seabrook 
Station UFSAR analyses.
    For the pressure and differential pressure sensors, calibration 
tests will detect degradation, which might significantly affect 
sensor response time.
    For the Process Protection racks, the Nuclear Instrumentation, 
and the Logic Systems calibration tests will continue to be 
performed which would detect significant degradation which might 
cause the response time to exceed the total allowance. The total 
time response allowance for each function bounds degradation that 
cannot be detected by the periodic surveillance.
    Thus, it is concluded that these proposed revisions to TS 3/
4.3.1 and TS 3/4.3.2 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 14, 2001.
    Description of amendment request: The proposed amendment will 
remove requirements for having the equipment hatch closed with four (4) 
bolts and one door of the personnel access lock (PAL) closed during 
core alterations and refueling operations for the Fort Calhoun Station 
(FCS). The technical specification (TS) for other containment 
penetrations will be modified to delete the requirement to be closed by 
an operable ventilation isolation actuation signal during core 
alterations and refueling operations. The proposed amendment will 
modify requirements for radiation monitors during core alterations and 
refueling operations. The TS Bases that are affected by the changes 
described above will be modified.
    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 changes to FCS TS modify requirements to have 
containment closure in place during core alterations and refueling 
operations in containment. These TS changes do not impact operation 
of other equipment or systems important to safety. The proposed TS 
changes reflect the parameters used in the radiological consequence 
calculations described in Section 5.0 of this license amendment 
request.
    The proposed change to TS 2.8.2(1) will be to delete the 
requirement for having equipment hatch closed and held in place by 
at least four (4) bolts and the requirement to have at least one 
door in the PAL closed. The requirements for containment penetration 
isolation via an operable VIAS [ventilation isolation actuation 
signal] have been deleted with these proposed changes. 
Administrative controls will be put in place instead for ``defense 
in depth'' action in regards to containment penetrations. These 
administrative controls include:
    a. The Equipment Hatch Enclosure (Room 66) doors or the 
equipment hatch and one door in the PAL shall be capable of being 
closed in less than one hour of a FHA [fuel handling accident].
    b. The Equipment Hatch Enclosure (Room 66) doors or the 
equipment hatch and one door in the PAL shall not be obstructed 
unless capability for rapid removal of obstructions is provided 
(such as quick disconnects for hoses).
    c. Penetrations providing direct access from the containment 
atmosphere to the outside atmosphere shall be capable of being 
closed on one side in less than one hour of a FHA.
    d. An individual or individuals shall be designated and 
available during core alterations and refueling operations, capable 
of closing the Equipment Hatch Enclosure (Room 66) doors or the 
equipment hatch, one door in the PAL, and penetrations that provide 
direct access from the containment atmosphere to the outside 
atmosphere.
    In addition, allowance will be granted to have penetration flow 
paths with direct access from the containment atmosphere to the 
outside atmosphere to be unisolated during core alterations and 
refueling operations. These proposed changes are based on a re-
analysis that was performed with respect to radiological 
consequences. The FHA re-analysis (Reference 10.1 [in the December 
14, 2001, submittal]) was performed in accordance with current 
accepted methodology, and consequences were expressed in TEDE [total 
effective dose equivalent] dose.
    The proposed change to TS 2.8.2(3) will delete the requirement 
for two gaseous radiation monitors being operable and supplied by 
independent power supplies. Instead, only one gaseous radiation 
monitor is required to be operable. VIAS actuation upon radiation 
monitor alert is not credited in the FHA re-analysis. VIAS actuation 
for containment purge or other penetration isolation is not 
credited.
    The current methodology as described in 10 CFR 50.67 specifies 
dose acceptance criteria in terms of TEDE dose. The revised FHA 
analysis results as discussed in Section 5.0 meet the applicable 
TEDE dose acceptance criteria (specified also in RG [regulatory 
guide] 1.183) for AST [alternative source term]. The most current 
FHA analysis does not credit containment integrity and, hence, is 
conservative in that aspect. These administrative controls proposed 
as stated above ensure that in the event of a FHA in containment 
(even though the containment fission product control function is not 
required to meet dose consequence criteria) that the Equipment Hatch 
Enclosure (Room 66) doors or the equipment hatch, one PAL door, and 
other pathways can be promptly closed.
    Currently the equipment hatch is closed with four (4) bolts, at 
least one PAL door closed, and other penetrations either are closed 
or capable of being closed on VIAS during core alterations and 
refueling operations to prevent the escape of radioactive material 
in the event of a FHA in containment. Whether the equipment hatch or 
other penetrations are open or closed during core alterations and 
refueling operations has no effect on the probability of any 
accident previously evaluated.
    Based on the TS changes approved in Reference 10.1, the changes 
being proposed in this amendment request will not affect assumptions 
contained in other plant safety analyses (Updated Safety Analysis 
Report) or the physical design of the plant, nor do they affect 
other TS that preserve safety assumptions.
    Therefore, the proposed changes do not involve a significant 
increase in the

[[Page 2927]]

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 current FHA analysis (Reference 10.1) assumes that all the 
iodine and noble gases become airborne, escape, and reach the site 
boundary and low population zone with no credit for filtration, 
containment closure, or deposition. Since the proposed changes do 
not involve the addition or modification of equipment nor alter the 
design of plant systems, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The changes proposed do not change how design 
basis accident (DBA) events were postulated nor do the changes 
themselves initiate a new kind of accident or failure mode with a 
unique set of conditions (proposed administrative controls). The FHA 
analysis documented in Reference 10.1 was performed consistent with 
10 CFR 50.67 and RG 1.183. Not crediting filtration systems for EAB/
LPZ dose consequences and only crediting natural forces is 
conservative from the aspect of dose consequences.
    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 implementation of the proposed changes does not reduce the 
margin of safety as defined in the alternate source term design 
basis site boundary and control room dose analyses (Reference 10.1). 
The radiological analyses results, with the proposed changes, remain 
within the regulatory acceptance criteria (10 CFR 50.67) utilizing 
the TEDE dose acceptance criteria directed in RG 1.183. These 
criteria have been developed for application to analyses performed 
with alternative source terms. These acceptance criteria have been 
developed for the purpose of use in design basis accident analyses 
such that meeting these limits demonstrates adequate protection of 
public health and safety. An acceptable margin of safety is inherent 
in these licensing limits.
    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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 14, 2001.
    Description of amendment request: The licensee proposes to (1) 
revise Technical Specifications 3.7(2)d and 3.7(4) to allow the tests 
to be performed on a refueling frequency outside of a refueling outage, 
and (2) correct the docket concerning inconsistencies in the 1973 Fort 
Calhoun Station (FCS) Safety Evaluation Report (SER) associated with 
the 13.8 kV transmission line capability.
    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 changes to Technical Specifications Sections 
3.7(2)d and 3.7(4) only provide greater flexibility in the time of 
testing. The periodicity remains the same, i.e., refueling 
frequency. There are no physical alterations proposed or being made 
to the D.C. emergency transfer switches or the 13.8 kV-480 V 
service. The proposed changes continue to address and comply with 
the regulatory requirements as described in Fort Calhoun Station 
Responses to 70 Criteria, Reference 9.2. The proposed changes will 
continue to assure that the D.C. emergency transfer switches and the 
13.8 kV-480 V service will perform their design function. Therefore, 
the proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes will not result in any physical alterations 
to the D.C. emergency transfer switches or 13.8 kV-480 V service, or 
any plant configuration, systems, equipment, or operational 
characteristics. There will be no change in operating modes or 
safety limits. With the proposed changes, the technical 
specifications retain requirements for operability and functionality 
on a refueling frequency. Therefore, the proposed changes do 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 changes provide flexibility in the time of 
performance of the required surveillance tests. The proposed changes 
will not alter any physical or operational characteristics of the 
D.C. emergency transfer switches or the 13.8 kV-480 V service. The 
proposed surveillance requirements will continue to assure that the 
design functions are met. Therefore, the proposed changes do 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 14, 2001.
    Description of amendment request: The licensee has proposed to 
change Technical Specification (TS) 2.10.4, to decrease the minimum 
required reactor coolant system (RCS) flow rate from 206,000 gallons 
per minute (gpm) to 202,500 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, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment to the RCS flow rate is the same as the 
indicated RCS flow rate prior to the TS Amendment 193 (Reference 
10.1 [in the December 14, 2001, submittal]). The plant was operated 
with the same RCS flow rate as the proposed value prior to Amendment 
193. Chapter 14 events and design basis accidents were analyzed with 
the RCS flow rate of 202,500 gpm using NRC approved methodology.
    In 1999 Fort Calhoun Station was granted TS Amendment 193 to 
increase the minimum indicated RCS flow rate to 206,000 gpm as a 
result of the removal of the steam generator orifice plates. 
Transient and thermal hydraulic analyses were performed using the 
amended RCS flow rate to verify that the minimum departure from 
nucleate boiling ratio (MDNBR) does not fall below the limiting 
value that supports the DNB specified acceptable fuel design limits.
    The FRA-ANP analysis confirms that the proposed reduction in RCS 
flow rate does not degrade the margin to the mechanical fuel design 
limits and that the fuel design criteria continue to be met.
    In view of the above confirmation, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to the RCS flow rate is not new since the 
plant was operating with

[[Page 2928]]

the same value prior to TS Amendment 193. The proposed revision does 
not change any equipment required to mitigate the consequences of an 
accident. OPPD will continue to analyze all applicable USAR Chapter 
14 events and design basis accidents as part of the reload analyses 
to establish the safety margin to the mechanical fuel design limits 
and confirm that all the fuel design criteria continue to be met. 
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 decreased RCS flow rate has been analyzed for thermal 
hydraulic effects on the reactor core. The analysis has confirmed 
that the proposed amendment does not degrade the margin to the 
mechanical fuel design limits and meets the fuel design criteria. 
The RCS flow rate surveillance requirements will continue to assure 
that the design functions are met. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.

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

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 14, 2001.
    Description of amendment request: The proposed amendment deletes 
technical specification (TS) Figures 
2-1A (Reactor Coolant System (RCS) Pressure--Temperature Limits for 
Heatup) and 2-1B (RCS Pressure--Temperature Limits for Cooldown) and 
replaces them with the single TS Figure 2-1. Additionally, the licensee 
proposes to change the lowest service temperature from 182 deg.F to 
164 deg.F to be in compliance with Reference 4, American Society of 
Mechanical Engineers (ASME) Section III, NB-2332 and the basis for the 
minimum boltup temperature to be in compliance with Reference 5, ASME 
Section XI, Appendix G. The Basis section for Technical Specification 
2.1.2 is being updated to reflect the use of ASME Code Case N-640 and 
Westinghouse Electric Company/Combustion Engineering's (W/CE) pressure 
temperature (P-T) limit curve methodology as applicable. Finally, based 
on the replacement of Figures 2-1A and 2-1B with the single Figure 2-1, 
the following TS are required to be changed: 2.1.1(8), 2.1.2, 2.1.2(1), 
2.1.2(2), 2.1.2(6), 2.1.2(6)(a), 2.1.2(6)(c), 2.1.2(6)(d), and 2.1.6(4) 
as they reference the deleted curves.
    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 changes will not increase the probability or 
consequence of any accident for the following reasons:
    (1) TS Figure 2-1 is proposed to incorporate the use of ASME 
Code Case N-640, which has been approved by the NRC as being 
acceptable for the development of P-T curves. Additionally, it is 
being updated for operation to higher neutron fluence values for use 
in the ART [adjacent reference temperature] calculations.
    (2) Reducing the lowest service temperature is in compliance 
with Reference 10.9, Section III, NB-2332.
    (3) The shift in the basis for minimum boltup temperature is in 
compliance with 10 CFR 50 Appendix G.
    (4) Updating the fluence and EFPY [effective full power years] 
applicability is in compliance with Regulatory Guide 1.99, Revision 
2.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed revision does not change any equipment required to 
mitigate the consequences of an accident. The continued use of the 
same Technical Specification administrative controls prevents the 
possibility of a new or different kind of accident. Since the 
proposed changes do not involve the addition or modification of 
equipment nor alter the design of plant systems, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The changes 
proposed do not change how design basis accident events are 
postulated nor do the changes themselves initiate a new kind of 
accident or failure mode with a unique set of conditions (proposed 
administrative controls). 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 Figure 2-1 does not constitute a significant 
reduction in the margin of safety due to the following:
    (1) The current LTOP [low temperature overpressure protection] 
analysis setpoints are bounding and applicable to this TS Figure.
    (2) The use of ASME Code Case N-640 has been approved by the NRC 
as acceptable for the development of P-T limit curves. W/CE's P-T 
limit curve methodology has been approved for the development of P-T 
curves.
    (3) The reduction in lowest service temperature is in compliance 
with Reference 10.9, Section III, NB-2332.
    (4) The shift in the basis of the minimum boltup temperature 
from NDTT to RTNDT is in compliance with Reference 10.4, 
Section XI, Appendix G.
    (5) Updating the fluence and EFPY applicability of the TS Figure 
2-1 to maintain validity is in compliance with Regulatory Guide 
1.99, Revision 2.
    The P-T curve results, with the proposed changes, remain within 
the regulatory acceptance criteria utilizing W/CE methodology and 
ASME Code Case N-640. These criteria, 10 CFR 50.36(c)(2), have been 
developed for application to analyses performed for long term 
operation of reactor vessels. These acceptance criteria have been 
developed for the purpose of use in design basis accident analyses 
such that meeting these limits demonstrates adequate protection of 
public health and safety. An acceptable margin of safety is inherent 
in these licensing limits. Therefore, the proposed changes do 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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 amendment request: December 14, 2001.
    Description of amendment request: A change is proposed to 
Surveillance Requirement (SR) 3.0.3 to allow a longer period of time to 
perform a missed surveillance. The time is extended from the current 
limit of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * *up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3 ``A risk 
evaluation shall be performed for any surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model

[[Page 2929]]

safety evaluation and model no significant hazards consideration (NSHC) 
determination, using the consolidated line item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated December 14, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: August 2, 2001.
    Description of amendment request: Proposed amendments would revise 
Technical Specification 3/4.6.1.6, ``Containment Structural 
Integrity,'' and replace it with reference to containment Post-
Tensioning System Surveillance Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.91, this analysis provides a determination 
that the proposed changes to the Technical Specifications do not 
involve any significant hazards consideration as defined in 10 CFR 
50.92.
    Criterion 1: Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes revise the surveillance requirements for 
the containment post-tensioning inservice inspection program as 
required by 10 CFR 50.55a(b)(2)(vi) and 10 CFR 50.55a(b)(2)(viii). 
The revised requirements do not affect the function of the 
containment post-tensioning system components. The post-tensioning 
systems are passive components whose failure modes could not act as 
accident initiators or precursors.
    The proposed changes do not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, this 
proposed change does not represent a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2: Does the proposed change create the possibility of 
a new or different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. The function of the containment 
post-tensioning system components are not altered by this change. 
Additionally, there is no change in the types or increases in the 
amounts of any effluent that may be released off-site and there is 
no increase in individual or cumulative occupational exposure. 
Therefore, this proposed change does not create the possibility of 
an accident of a different kind than previously evaluated.
    Criterion 3: Does the proposed change involve a significant 
reduction in the margin of safety?
    Response: No.
    The proposed change does not impact the margin of safety 
included in the design pressure compared to the peak calculated 
pressure because the proposed activity does not alter, in any way, 
the available force provided by the tendons. Therefore, this 
proposed change does not involve a significant reduction in a margin 
of safety.
    Based on the evaluation provided above, the proposed changes do 
not involve a significant hazards consideration under 10 CFR 
50.92(c), and will not have a significant effect on the safe 
operation of the plant. Therefore, there is reasonable assurance 
that operation of the South Texas Project in accordance with the 
proposed revised Technical Specifications will not endanger the 
public health and safety.

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

[[Page 2930]]

10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the request for amendments involves no significant 
hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 22, 2001.
    Description of amendments request: The proposed amendments would 
revise the Technical Specification Limiting Condition for Operation for 
Containment Penetrations to allow the equipment hatch to be open during 
core alterations and/or during movement of irradiated fuel assemblies 
within containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    STPNOC [South Texas Project Nuclear Operating Company] has 
evaluated whether the proposed amendment involves a significant 
hazards consideration by focusing on the three standards set forth 
in 10 CFR 50.92 as discussed below:
    (1) Will operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes will allow the equipment hatch to be open 
during core alterations and movement of irradiated fuel assemblies 
inside containment. The status of the equipment hatch during 
refueling operations has no affect on the probability of the 
occurrence of any accident previously evaluated. The proposed 
revision does not alter any plant equipment or operating practices 
in such a manner that the probability of an accident is increased. 
Since the consequences of a fuel handling accident inside 
containment with an open equipment hatch are bounded by the current 
analysis described in the UFSAR [Updated Final Safety Analysis 
Report] and the probability of an accident is not affected by the 
status of the equipment hatch, the proposed change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not create any new failure modes for any 
system or component, nor do they adversely affect plant operation. 
No new equipment will be added and no new limiting single failures 
will be created. The plant will continue to be operated within the 
envelope of the existing safety analyses. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident previously evaluated.
    (3) Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The previously determined radiological dose consequences for a 
fuel handling accident inside containment with the personnel airlock 
doors open remain bounding for the proposed changes. These 
previously determined dose consequences were determined to be well 
within the limits of 10 CFR 100 and they meet the acceptance 
criteria of SRP section 15.7.4 and GDC 19. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 3, 2001.
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TSs) for the Auxiliary Feedwater 
(AFW) System to provide consistent allowed outage times (AOT) and 
required actions for any inoperable motor driven AFW pump(s). The AOT 
for one inoperable motor drive AFW pump is also proposed to be extended 
from 72 hours to 28 days based on a risk-informed approach.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    STPNOC [South Texas Project Nuclear Operating Company] has 
evaluated whether a significant hazards consideration is involved 
with the proposed amendment by focusing on the three standards set 
forth in 10 CFR 50.92 as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS change reflects the STP four-train AFWS design 
in the required actions and AOTs. No actual plant equipment or 
accident analyses will be affected by the proposed change. 
Therefore, the proposed AOT change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The administrative change of deleting the words ``At least'' 
clarifies that there are only four AFW pumps in the design. The 
administrative change involves no increase in the probability or 
consequences of an accident.
    If all four AFW trains are inoperable in Mode 1, 2, or 3, the 
unit is in a seriously degraded condition with only limited means 
for conducting a cooldown. In such a condition, the unit should not 
be perturbed by any action, including a power change that might 
result in a trip. The seriousness of this condition requires that 
action be started immediately to restore one AFW train to operable 
status. Required Action (d) is modified by adding a sentence 
indicating that all required mode changes or power reductions are 
suspended until one AFW train is restored to operable status. This 
statement reflects the same sentence for the case of all AFW trains 
being inoperable in NUREG-1431, TS 3.7.5. In this case, LCO 
[limiting condition for operation] 3.0.3 is not applicable because 
it could force the unit into a less safe condition. Therefore, the 
addition of the sentence to Action (d) involves no increase in the 
probability or consequences of an accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed TS change reflects the STP four-train AFWS design 
in the required actions and AOTs. No actual plant equipment or 
accident analyses will be affected by the proposed change and no 
failure modes not bounded by previously evaluated accidents will be 
created. Therefore, the proposed AOT change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The administrative change of deleting the words ``At least'' 
clarifies that there are only four AFW pumps in the design. The 
change does not create the possibility of any accident.
    Required Action (d) is modified by adding a sentence indicating 
that all required mode changes or power reductions are suspended 
until one AFW train is restored to operable status. This statement 
reflects the same sentence for the case of all AFW trains being 
inoperable in NUREG-1431, TS 3.7.5. In this case, LCO 3.0.3 is not 
applicable because it could force the unit into a less safe 
condition. Therefore, the addition of the sentence to Action (d) 
does 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?

[[Page 2931]]

    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, reactor 
coolant pressure boundary, and containment structure) to limit the 
level of radiation dose to the public.
    The proposed TS change reflects the STP four-train AFWS design 
in the required actions and AOTs. No actual plant equipment or 
accident analyses will be affected by the proposed change. 
Additionally, the proposed change will not relax any criteria used 
to establish safety limits, will not relax any safety systems 
settings, and will not relax the bases for any limiting conditions 
of operation. Therefore, the proposed AOT change does not involve a 
significant reduction in a margin of safety.
    The administrative change of deleting the words ``At least'' 
clarifies that there are only four AFW pumps in the design. The 
change does not involve any reduction in a margin of safety.
    Required Action (d) is modified by adding a sentence indicating 
that all required mode changes or power reductions are suspended 
until one AFW train is restored to operable status. This statement 
reflects the same sentence for the case of all AFW trains being 
inoperable in NUREG-1431, TS 3.7.5. In this case, LCO 3.0.3 is not 
applicable because it could force the unit into a less safe 
condition. Therefore, the addition of the sentence to Action (d) 
does not involve a significant reduction in a margin of safety.
    Based on the above, STPNOC concludes that the proposed amendment 
involves no significant hazards consideration under the standards 
set forth in 10 CFR 50.92, and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 18, 2001.
    Brief description of amendments: A change is proposed to 
Surveillance Requirement (SR) 3.0.3 to allow a longer period of time to 
perform a missed surveillance. The time is extended from the current 
limit of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on June 14, 2001 (66 FR 
32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the following NSHC 
determination in its application dated December 18, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

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

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: December 6, 2001.
    Description of amendment request: The amendment would revise the 
following Technical Specifications (TSs): (1) TS 3.3.6, ``Containment 
Purge Isolation Instrumentation;'' (2) TS 3.3.7, ``Control Room 
Emergency Ventilation

[[Page 2932]]

System (CREVS) Instrumentation;'' and (3) TS 3.9.4, ``Containment 
Penetrations.'' The revisions to TS 3.3.6 would alter Condition C of 
the Actions for the Limiting Condition for Operation (LCO), and delete 
footnotes (a) and (b) from applicable Modes for the automatic actuation 
logic and actuation relays function and the containment purge exhaust 
radiation gaseous function in Table 3.3.6-1. The revisions to TS 3.3.7 
would add (1) Surveillance Requirement (SR) 3.3.7.6, (2) footnote (c) 
to Table 3.3.7-1, (3) the fuel building exhaust radiation gaseous 
function to the table, and (4) footnote (c), 2 trains, and SR 3.3.7.6 
to the applicable Modes, required channels, and surveillance 
requirements columns for the automatic actuation logic and actuation 
relays and control room radiation control room air intakes functions in 
Table 3.3.7-1. The revisions to TS 3.9.4 are to add the phrase ``or if 
open, capable of being closed'' to item a on the equipment hatch in LCO 
3.9.4, delete the word ``closed'' from item b on the emergency air lock 
in LCO 3.9.4, add SR 3.9.4.2 on verifying the capability to install the 
equipment hatch when it is open, and renumber the existing SR 3.9.4.2 
to SR 3.9.4.3. The revisions to the TSs are to allow the equipment 
hatch and the emergency air lock to be open in refueling outages during 
core alterations and/or movement of irradiated fuel within containment. 
The revisions to TSs 3.3.6 and 3.3.7 are to eliminate the requirement 
for automatic actuation of containment purge isolation during core 
alterations and/or during movement of irradiated fuel to allow the 
containment purge system to remain in operation during refueling when 
the equipment hatch is open, and to add a new surveillance to response 
time test the channels for the control room radiation monitor 
detectors, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes will allow the equipment hatch [or the 
emergency air lock] to be open during CORE ALTERATIONS and movement 
of irradiated fuel assemblies inside containment. The status of the 
equipment hatch or the emergency air lock during refueling 
operations has no affect on the probability of the occurrence of any 
accident previously evaluated. The proposed revision does not alter 
any plant equipment or operating practices in such a manner that the 
probability of an accident is increased. Since the consequences of a 
fuel handling accident inside containment with an open equipment 
hatch [or open emergency air lock] are bounded by the current 
analysis described in the FSAR [Final Safety Analysis Report] and 
the probability of an accident is not affected by the status of the 
equipment hatch [or the emergency air lock], the proposed change[s 
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 changes do not create any new failure modes for any 
system or component, nor do they adversely affect plant operation. 
No new equipment will be added and no new limiting single failures 
will be created. The plant will continue to be operated within the 
envelope of the existing safety analysis.
    Therefore, the proposed changes do not create a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The previously determined radiological dose consequences for a 
fuel handling accident inside containment with the [equipment hatch 
or emergency] air lock doors open remain bounding for the proposed 
changes. Those previously determined dose consequences were 
determined to be well within the limits of 10 CFR 100 and they meet 
the acceptance criteria of SRP [Standard Review Plan] section 15.7.4 
and GDC 19 [for exposure of control room operators].
    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: John O'Neill, Esq., Shaw, Pittman, Potts & 
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.

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

    Date of application for amendments: September 21, 2001.
    Brief description of amendments: Will revise the Technical 
Specifications to allow Sequoyah to insert tritium-producing burnable 
absorber rods into the reactor core.
    Date of publication of individual notice in the Federal Register: 
December 17, 2001 (66 FR 65000).
    Expiration date of individual notice: January 16, 2002.

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

    Date of application for amendments: April 20, 2001.
    Brief description of amendments: Will revise the Final Safety 
Analysis Report to reflect a change in the spent fuel pool cooling 
analysis methodology.
    Date of publication of individual notice in the Federal Register: 
December 17, 2001 (66 FR 64998).
    Expiration date of individual notice: January 16, 2002.

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

    Date of application for amendments: August 20, 2001.
    Brief description of amendments: Will revise the Technical 
Specifications to allow Watts Bar to insert tritium-producing burnable 
absorber rods into the reactor core.
    Date of publication of individual notice in the Federal Register: 
December 17, 2001 (66 FR 65005).
    Expiration date of individual notice: January 16, 2002.

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

[[Page 2933]]

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/reading-rm/adams.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 e-mail to [email protected]. 

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: May 31, 2001, as supplemented 
August 1, 2001, and September 26, 2001.
    Brief description of amendment: The amendment approves a change to 
the Technical Specifications and Bases associated with the operability 
of A.C. electrical power sources to increase the allowed outage time 
(AOT) for one inoperable emergency diesel generator (EDG) from 72 hours 
to 14 days. This change to the AOT allows the performance of various 
EDG maintenance and repair activities during plant operation.
    Date of issuance: January 4, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 261.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.

Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41614).

    The August 1, 2001, and September 26, 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.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41614). The August 1, 2001, and September 26, 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 January 4, 2002.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: July 31, 2001.
    Brief description of amendment: The amendment deletes Technical 
Specifications Section 6.18, ``PASS [Post Accident Sampling System]/
Sampling and Analysis of Plant Effluents,'' for Millstone Nuclear Power 
Station, Unit No. 2 and thereby eliminates the requirements to have and 
maintain the post-accident sampling program.
    Date of issuance: January 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 262.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55009).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 31, 2001.
    Brief description of amendment: The amendment deletes Technical 
Specifications Section 6.8.4.d, ``Post Accident Sampling,'' for 
Millstone Nuclear Power Station, Unit No. 3 and thereby eliminates the 
requirements to have and maintain the post-accident sampling program.
    Date of issuance: January 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 201.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55011).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 23, 2001, as supplemented 
by letters dated June 29, July 19, and November 13, 2001.
    Brief Description of amendment: The amendment changes Technical 
Specification (TS) 3.4.1.6, ``Reactor Coolant System--Isolated Loop 
Startup'' which includes revisions to the limiting condition for 
operation. Some of the changes to TS 3.4.1.6 affect restrictions that 
were included as part of the original Millstone Unit 3 licensing basis 
allowing power operation with one isolated reactor coolant system loop.
    Date of issuance: January 9, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 202.
    Facility Operating License No. NF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36338). The letters dated June 29, July 19, and November 13, 2001, 
provided clarifying information and did not change the staff's initial 
proposed no significant hazards consideration determination or

[[Page 2934]]

expand the scope of the application as published in the Federal 
Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 9, 2002.
    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: April 13, 2001.
    Brief description of amendment: The change relaxes the allowable 
cooldown rate in the Reactor Coolant System (RCS) Technical 
Specifications (TS) 3.4.8.1, ``Pressure/Temperature Limits.'' 
Specifically, the change eliminates the limitation of a 10  deg.F per 
hour cooldown rate when the RCS temperature is below 135  deg.F. The 
proposed limitations permit a 100 oF per hour cooldown rate to continue 
down to an RCS temperature of 110  deg.F, at which point the rate is 
reduced to 30  deg.F per hour.
    Date of issuance: January 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 177.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22029).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 8, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: July 9, 2001.
    Brief description of amendments: The proposed amendments would 
incorporate TS changes that are being made to provide consistency with 
the changes to 10 CFR 50.59, ``Changes, tests, and experiments,'' as 
published in the Federal Register (64 FR 53582), dated October 4, 1999. 
Specifically, the changes replace the terms ``safety evaluation'' with 
``10 CFR 50.59 evaluation'' and ``unreviewed safety question'' with 
``requires NRC approval pursuant to 10 CFR 50.59.''
    Date of issuance: December 28, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 125 and 120.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44170).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: September 21, 2001.
    Brief description of amendments: The amendments delete Technical 
Specification 5.5.3, ``Post Accident Sampling,'' and thereby eliminate 
the requirement to have and maintain the Post Accident Sampling System 
at the Braidwood Station, Unit Nos. 1 and 2, and Byron Station, Units 
Nos. 1 and 2.
    Date of issuance: December 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 365 days.
    Amendment Nos.: 126 and 121.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55018).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 15, 2001, as supplemented 
by letter dated November 12, 2001.
    Brief description of amendments: The proposed amendments would 
allow use of ATRIUM 10 fuel from Framatome Advanced Nuclear Fuel, Inc.
    Date of issuance: December 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 152 and 138.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41618).
    The November 12, 2001, submittal was clarifying in nature and did 
not change the scope of the original notice or proposed no significant 
hazards finding dated August 8, 2001 (66 FR 41618). The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated December 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 14, 2000, as supplemented 
December 19, 2001.
    Brief description of amendment: The amendment changes the operating 
license to reflect a change in the name of IES Utilities, Inc., a co-
owner of the Duane Arnold Energy Center and licensee, to Interstate 
Power and Light Company.
    Date of issuance: January 2, 2002.
    Effective date: As of January 1, 2002, and shall be implemented 
within 30 days.
    Amendment No.: 244.
    Facility Operating License No. DPR-49: The amendment revised the 
operating license.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46009).
    The December 19, 2001, supplemental letter provided notification 
that (1) the required regulatory approvals for the merger had been 
received and (2) the projected schedule for the merger was January 1, 
2002. The supplemental letter did not change the staff's initial 
proposed no significant hazards consideration determination or expand 
the application beyond the scope of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 2, 2002.
    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: January 8, 2001, as supplemented 
on February 6, December 7, and December 27, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.5.1.b.1 to change the minimum acceptable Core 
Spray subsystem flow from 6,350 gallons per minute (gpm) to 6,150 gpm.
    Date of issuance: January 7, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.

[[Page 2935]]

    Amendment No.: 136.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 22, 2001 (66 FR 
6701).
    The letters dated February 6, December 7, and December 27, 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 January 7, 2002.
    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: April 16, 2001, as supplemented 
on July 5, 2001.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) requirements associated with the operation and 
surveillance testing of the 28 Volt D.C. (VDC) Batteries. The revised 
Limiting Condition for Operation (LCO) and Surveillance Requirements 
(SRs) are now more consistent with the 125 VDC Battery System LCO and 
SRs as well as similar to standard TSs provided by NUREG-1431, 
``Standard Technical Specifications, Westinghouse Plants,'' Revision 1, 
dated April 1995.
    Date of issuance: January 4, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 249 and 229.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57124).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 4, 2002.
    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: June 19, 2001, as supplemented by 
letters dated August 15, August 31, November 20, and December 17, 2001.
    Brief description of amendments: The amendments modified Facility 
Operating License Nos. NPF-87 and NPF-89 to reflect the direct transfer 
of control of TXU Electric Company's operating authority and 100-
percent ownership interest in the Comanche Peak Steam Electric Station, 
Unit Nos. 1 and 2, to a newly formed generating company: TXU Generation 
Company LP.
    Date of issuance: January 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 7 days from the date of issuance.
    Amendment Nos.: 90 and 90.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: August 20, 2001 (66 FR 
43594).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 21, 2001.

    Dated at Rockville, Maryland, this 11th day of January 2002.

    For the Nuclear Regulatory Commission,
John A. Zwolinski,
Director, Division of Licensing, Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-1210 Filed 1-18-02; 8:45 am]
BILLING CODE 7590-01-P