[Federal Register Volume 67, Number 24 (Tuesday, February 5, 2002)]
[Notices]
[Pages 5323-5347]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-2567]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 11, 2002 through January 24, 2002. 
The last biweekly notice was published on January 22, 2002 (67 FR 
2917).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the

[[Page 5324]]

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 March 7, 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 5325]]

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 lower the 
maximum allowable differential pressure across the Engineered Safety 
Features (ESF) ventilation system units when tested at specified system 
flowrates.
    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.
    Technical Specification (TS) 5.5.11, Ventilation Filter Testing 
Program (VFTP) establishes a program for requiring testing of 
Engineered Safety Feature (ESF) filter ventilation systems in 
accordance with appropriate regulatory guidance.
    PVNGS [Palo Verde Nuclear Generating Station] calculations 13-
MC-HJ-0804 and 13-MC-HF-0902 were developed to document the design 
basis and testing standard positions that PVNGS has taken concerning 
the Control Room Essential Filtration System (CREFS) air filtration 
units (AFUs) and the ESF Pump Room Exhaust Air Cleanup System 
(PREACS) AFUs. These calculations established a lower design dirty 
filter differential pressure (D/P) to ensure that the AFUs are 
capable of delivering the design flows at 100% maximum dirty filter 
condition and also able to meet the adsorber residence time when the 
filters are clean. Design margin of the AFUs is validated via 
analyses performed in the referenced calculations and confirmed by 
the various startup and surveillance tests.
    The analyses established a more restrictive design criteria than 
that which is currently listed in TS 5.5.11.d. The new D/P limit for 
the CREFS AFUs is less than or equal to 4.8 inches water gauge 
(iwg). The new D/P limit for the PREACS AFUs is less than or equal 
to 5.2 iwg. This applies to all three of the PVNGS units. Each PVNGS 
unit is equipped with two CREFS and two PREACS AFUs.
    These essential AFUs are not event initiators. The essential 
CREFS and PREACS AFUs are used to mitigate the consequences of a 
postulated accident as discussed in Updated Final Safety Analysis 
Report (UFSAR) Sections 15.6 and 15.7. The proposed change in filter 
D/P for dirty filter conditions does not increase the probability of 
an accident previously evaluated.
    The accident analyses that could be affected by the proposed 
changes to the CREFS and PREACS AFUs are addressed in the 
calculations which determine the expected radiological doses in the 
control room, at the Exclusion Area Boundary (EAB), and in the Low 
Population Zone (LPZ) resulting from postulated accidents. The 
efficiency of the essential AFU filter and charcoal adsorber as well 
as adsorber residence time and airflow rate are required parameters 
to evaluate the removal of radioactive gases and particulates from 
the postulated accidents evaluated in UFSAR Chapter 15. However, the 
proposed changes to the essential AFUs D/P limits ensure that PVNGS 
remains within existing licensing bases for radiological 
consequences of fuel handling accidents and LOCA events.
    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 purpose of the essential AFUs (CREFS and PREACS) is to 
mitigate the consequences of an accident and as such, they are not 
plant accident initiators.
    The proposed changes in filter D/P limits for these essential 
AFUs do not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operations. The proposed changes in 
the filter D/P limit for dirty filter conditions ensure that PVNGS 
remains within existing licensing bases for radiological 
consequences of fuel handling accidents and LOCA [loss-of-coolant 
accident] events and are not initiators of any new or different 
kinds of accidents.
    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 change in the allowed maximum D/P across the filter 
in a dirty condition is a more conservative and restrictive change 
(less than or equal to 4.8 inches of water (iwg) for the CREFS units 
and 5.2 iwg for the PREACS units) than the current value of ``less 
than 8.4 iwg'' in Technical Specification 5.5.11.d. Under these 
conditions, the AFUs are required to deliver the design flows at a 
lower maximum D/P, which increases the structural safety margin of 
the filters. At the same time, the charcoal adsorber residence time 
requirements are met for the higher fan flowrate achieved with clean 
filters. The variations in diesel generator output voltage and 
frequency and its effects on the airflows and adsorber residence 
time are bounded by the design value parameters as demonstrated in 
calculations 13-MC-HJ-0804 and 13-MC-HF-0902. As such, the proposed 
changes ensure that PVNGS remains within existing licensing bases.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
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.

Carolina Power & Light Company (CP&L), Docket No. 50-261, H. B. 
Robinson Steam Electric Plant, Unit No. 2, Darlington County, South 
Carolina

    Date of amendment request: December 20, 2001.
    Description of amendment request: The amendment would revise 
Technical Specifications Section 5.6.5, ``Core Operating Limits Report 
(COLR)'' to add a report to the list of documents describing the 
approved methodologies.
    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 report proposed to be added to the COLR references is 
under generic review by NRC and, if approved, will be adopted for 
use.] Analyzed events are assumed to be initiated by the failure of 
plant structures, systems, or components. The core operating limits 
developed in accordance with the new methodology will be bounded by 
any limitations in the NRC acceptance in its safety evaluations of 
the new methodologies. The topical report associated with the new 
methodology demonstrates that the integrity of the fuel will be 
maintained during normal operations and that design requirements 
will continue to be met. The proposed change does not involve 
physical changes to any plant structure, system, or component. 
Therefore, the probability of occurrence for a previously analyzed 
accident is not significantly increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. The proposed methodology continues to meet 
applicable design and safety analyses acceptance criteria. The 
proposed change does not affect the performance of any equipment 
used to mitigate the consequences of an analyzed accident. As a 
result, no analysis assumptions are violated and there are no 
adverse effects on the factors that contribute to offsite or onsite 
dose as the result of an accident. The proposed change does not 
affect setpoints that initiate protective or

[[Page 5326]]

mitigative actions. The proposed change ensures that plant 
structures, systems, or components are maintained consistent with 
the safety analysis and licensing bases. Based on this evaluation, 
there is no significant increase in the consequences of a previously 
analyzed event.
    Therefore, this 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 Previously Evaluated The 
proposed change does not involve any physical alteration of plant 
systems, structures, or components, other than allowing for fuel 
design in accordance with NRC approved methodologies. The proposed 
methodology continues to meet applicable criteria for LBLOCA [large-
break loss-of-coolant accident] analysis. No new or different 
equipment is being installed. No installed equipment is being 
operated in a different manner. There is no alteration to the 
parameters within which the plant is normally operated or in the 
setpoints that initiate protective or mitigative actions. As a 
result no new failure modes are being introduced. There are no 
changes in the methods governing normal plant operation, nor are the 
methods utilized to respond to plant transients altered. 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 the Margin of Safety
    The margin of safety is established through the design of the 
plant structures, systems, and components, through the parameters 
within which the plant is operated, through the establishment of the 
setpoints for the actuation of equipment relied upon to respond to 
an event, and through margins contained within the safety analyses. 
The proposed change in the methodology used for LBLOCA analyses does 
not impact the condition or performance of structures, systems, 
setpoints, and components relied upon for accident mitigation. The 
proposed change does not significantly impact any safety analysis 
assumptions or results. Therefore, the proposed change does not 
result in a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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 6, 2001.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) 3.7.16, ``Control Room Area 
Cooling System (CRACS),'' which currently requires entry into TS 3.0.3 
when two trains of CRACS are inoperable. The proposed amendments would 
allow 6 hours to restore the operability of one train.
    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.
    Loss of CRACS for the duration of the Completion Time is not a 
safety concern because equipment in the control area is suitable for 
considerably higher temperatures than will be experienced within the 
Completion Time.
    The accidents evaluated in the UFSAR [Updated Final Safety 
Analysis Report] are not initiated by the CRACS or loss of the 
CRACS. Furthermore, the CRACS is not directly credited for 
mitigation of the accidents evaluated in the UFSAR. The CRACS does 
perform a support function to maintain environmental conditions for 
equipment that does help mitigate accidents. The proposed change 
does extend the total time from loss of a second required train 
until entry into the required MODEs. However, analysis confirms that 
the CRACS function is not required for a number of hours (i.e. 18 or 
more), which is substantially greater than the proposed Completion 
Time of 6 hours. The proposed Completion Time of 6 hours allows 
reasonable time for restoration prior to initiation of shutdown 
while leaving sufficient time to reach hot shutdown. The probability 
of an accident or event occurring during this Completion Time is 
acceptably low.
    The current TS may require simultaneous reduction in power and 
shutdown of all three Units. Such action is not without some risk. 
Allowing the requested limited additional time to restore control 
area cooling reduces some risk factors by not changing plant power 
level in response to a minor problem that does not constitute a 
safety concern. If the initiation of shutdown of the affected units 
does become necessary, this change would allow operators more 
flexibility to sequence the shutdowns to minimize overall operator 
burden and the impact of simultaneous shutdowns.
    In summary, this change will not involve a significant increase 
in the probability or consequences of any previously evaluated 
accident.
    2. Do the proposed changes create the possibility of new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different kind of accident has been identified as a 
result of this Technical Specification change.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The accidents evaluated in the UFSAR are not initiated by the 
CRACS or loss of the CRACS. The loss of the CRACS was screened out 
of the Oconee PRA and is not modeled in the present Oconee PRA as 
either an initiating event or as a support system failure. 
Temperature transient analyses calculate the time to reach the 
limiting design temperature of required systems, structures, or 
components supported by CRACS. Current analyses show CRACS is not 
required to perform a support function for at least 18 hours.
    This 18 hour time is not used to calculate the consequences or 
impact on fission product barriers if CRACS is not restored. Instead 
this time is used to prioritize activities to restore CRACS and is 
substantially greater than the proposed 6 hour Completion Time. As 
discussed above, this allows reasonable time for restoration prior 
to initiation of shutdown, while leaving sufficient time to reach 
hot shutdown. Since either the CRACS function will be restored or 
the affected unit(s) will be shutdown, this change would not result 
in a change of, or challenge to, the design basis limit for a 
fission product barrier.
    This change does not involve a departure from a method of 
evaluation used for evaluating behavior or response of the facility 
or supported components.

    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 Technical Specification 5.6.5.b to eliminate the revision 
number and dates of the topical reports that contain the analytical 
methods used to determine the core operating limits. This proposed 
change is consistent with

[[Page 5327]]

TSTF (Technical Specification Task Force)-363.
    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 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 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard J. Laufer, Acting.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: December 3, 2001.
    Description of amendment request: Energy Northwest is requesting a 
revision to the technical specifications (TSs) and licensing and design 
bases to reflect the application of alternative source term 
methodology. The alternative source term analyses have been performed 
without crediting secondary containment during fuel handling accidents. 
As such, the proposed license amendment relaxes operability 
requirements during fuel handling and core alterations for: (1) 
secondary containment; (2) secondary containment isolation 
instrumentation; and (3) the standby gas treatment system. The 
alternative source term analyses have also been performed without 
crediting the main steam leakage control system; therefore, the 
licensing basis and the TS are being revised to reflect the proposed 
deactivation of the system. The license amendment request also 
addresses the establishment of secondary containment vacuum under 
adverse environmental conditions. In addition, the amendment request 
increases the allowed amount of unfiltered control room leakage into 
the control room.
    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 alternative source term does not affect the design or 
operation of the facility; rather, once the occurrence of an 
accident has been postulated, the new source term is an input to 
evaluate the consequence. The implementation of the alternative 
source term methodology has been evaluated in revisions to the 
analyses of the following limiting design basis accidents at 
Columbia Generating Station:
     Control Rod Drop Accident
     Fuel Handling Accident
     Main Steam Line Break Accident
     Loss of Coolant Accident
    Based upon the results of these analyses, it has been 
demonstrated that, with the requested changes, the dose consequences 
of these limiting events are within the regulatory guidance provided 
by the NRC for use with the alternative source term. This guidance 
is presented in 10 CFR 50.67 and associated Regulatory Guide 1.183, 
and Standard Review Plan Section 15.0.1.
    Requirements for secondary containment operability, secondary 
containment isolation valves, and the standby gas treatment system 
during fuel movement or core alterations are being eliminated. This 
is acceptable because, with the application of alternative source 
term methodology, secondary containment is not credited for the fuel 
handling accident. The licensing basis is being revised to reflect 
the proposed deactivation of the main steam leakage control system. 
This is acceptable because, with the application of alternative 
source term methodology, no credit is assumed for the system in the 
accident analyses.
    With regard to the Justification for Continued Operation 
regarding the establishment of secondary containment vacuum under 
adverse environmental conditions, the proposed changes to the 
secondary containment and standby gas treatment system Technical 
Specifications and application of alternative source term 
methodology ensures that secondary containment draw-down and bypass 
leakage are within the assumptions of the applicable safety 
analysis.
    With regard to the previously-identified Unreviewed Safety 
Question pertaining to increased unfiltered control room in-leakage 
into the control room envelope, application of alternative source 
term methodology has shown that in-leakage rates in excess of tested 
values would result in control room doses below the regulatory 
limit.
    Therefore, operation of Columbia Generating Station in 
accordance with the proposed amendment will 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 alternative source term does not affect the design, 
functional performance or operation of the facility. Similarly, it 
does not affect the design or operation of any structures, systems 
or components equipment or systems involved in the mitigation of any 
accidents, nor does it affect the design or operation of any 
component in the facility such that new equipment failure modes are 
created.
    Requirements for the main steam leakage control system are being 
deleted by this proposed amendment request. This is acceptable 
because the system no longer meets the criteria of 10 CFR 50.36. 
With the application of alternative source term methodology, no 
credit is assumed for the system in the accident analyses. 
Furthermore, since the main steam leakage control system is a 
mitigating system, it cannot create the possibility of an accident.
    Requirements for secondary containment operability, secondary 
containment isolation valves, and the standby gas treatment system 
during fuel movement or core alterations are being eliminated. This 
is also acceptable because, with the application of alternative 
source term methodology, secondary containment is not credited for 
the fuel handling accident.
    Therefore, the operation of Columbia Generating Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The changes proposed are associated with the implementation of a 
new licensing basis for Columbia Generating Station. Approval of the 
basis change from the original source term developed in accordance 
with TID-14844 to a new alternative source term as described in 
Regulatory Guide 1.183 is requested by this submittal. The results 
of the accident analyses revised in support of this submittal, and 
the requested Technical Specification changes, are subject to 
revised acceptance criteria. These analyses have been performed 
using conservative methodologies.

[[Page 5328]]

    Safety margins and analytical conservatisms have been evaluated 
and are satisfied. The analyzed events have been carefully selected 
and margin has been retained to ensure that the analyses adequately 
bound postulated event scenarios. The dose consequences of these 
limiting events are within the acceptance criteria also found in the 
latest regulatory guidance. This guidance is presented in 10 CFR 
50.67 and associated Regulatory Guide 1.183.
    The proposed changes can be made while still satisfying 
regulatory requirements and review criteria, with significant 
margin. The changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries, as well as the 
control room, are within the corresponding regulatory limits.
    Therefore, operation of Columbia Generating Station in 
accordance with 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: September 7, 2001 as revised December 
17, 2001. This notice supersedes (66 FR 52799) published on October 17, 
2001, which was based upon the licensee's application dated September 
7, 2001.
    Description of amendment request: The amendment would revise the 
Post Accident Monitoring Instrumentation Technical Specifications to 
ensure that licensee commitments to Regulatory Guide 1.97 are properly 
reflected.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    The proposed amendment involves rewording or reformatting of 
technical specification requirements regarding certain post accident 
monitoring instrumentation at Indian Point 3, to improve the 
usability of the specification. The proposed rewording of the 
required channels for core exit temperature adopts the wording from 
the Standard Technical Specifications, which is applicable to the 
Indian Point 3 design. New condition entry statements are added in 
Condition C as an alternate formatting method which replaces the 
existing approach of using notes in the instrumentation list in 
Table 3.3.3-1, for certain instrument channels. Similarly, combining 
two existing functions into one new function is an improved 
formatting method that eliminates the need for a note in the Table. 
None of these proposed changes affect the requirements established 
in the existing specification.
    Post accident monitoring instrumentation is a tool used by plant 
operators to conduct diagnostic activities outlined in plant 
emergency operating procedures. The presence or absence of this 
instrumentation does not influence accident initiators for accidents 
previously analyzed. Also, this instrumentation is not credited to 
support automatic responses for accident mitigating systems or 
equipment. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed amendment involves rewording or reformatting of 
technical specification requirements to improve the usability of the 
specification for certain post accident monitoring instrumentation 
at Indian Point 3. The proposed amendment does not involve any 
changes to plant equipment, setpoints, or the way in which the plant 
is operated. The proposed amendment maintains the existing 
requirements for post accident monitoring instrumentation using an 
improved presentation format. Therefore the proposed amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    Response:
    The proposed amendment involves rewording or reformatting of 
technical specification requirements to improve the usability of the 
specification for certain post accident monitoring instrumentation 
at Indian Point 3. The proposed rewording of the required channels 
for core exit temperature adopts the wording from the Standard 
Technical Specifications, which is applicable to Indian Point 3. Use 
of the standard wording ensures consistent application of the 
requirements for this post accident monitoring function. Similarly, 
reformatting the specification to use new condition entry 
statements, rather than the existing notations in the Table will 
improve the usability of the specification and ensure that the 
intended requirements will be consistently applied.
    The proposed changes do not delete or modify existing 
requirements or add new requirements. The changes involve rewording 
or reformatting of existing requirements and provide an improved 
method of stating the requirements intended in the existing 
specification. Therefore, the proposed amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Joel T. Munday, Acting.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: January 16, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to permit functional testing 
of the emergency diesel generators (EDGs) to be performed during power 
operation. The proposed changes will add a footnote to Surveillance 
Requirement 4.8.1.1.2.g.7 regarding the 24-hour functional test of the 
EDGs. The changes are based on an integrated review of deterministic 
design basis factors, and an evaluation of plant risk using 
probabilistic safety assessment (PSA) techniques.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The function of the emergency diesel generators is to supply 
emergency power in the event of a LOOP. Operation of the EDGs is not 
a precursor to any accident. The EDGs provide assistance in accident 
mitigation. There are no technical changes related to the acceptance 
criteria of the surveillance requirement. The proposed change 
requesting that the scheduling aspects of the surveillance 
requirements be changed to accommodate improved planning capability 
for testing does not affect the accident analyses. The EDG that is 
being tested will be considered inoperable however, the remaining 
required EDGs would be operable during the test and they are capable 
of supporting the safe shutdown of the plant. The Probabilistic 
Safety Assessment (PSA)

[[Page 5329]]

results fall below the Acceptance Guidelines for TS changes 
contained in Regulatory Guides 1.174 and 1.177; therefore, the risk 
of performing the EDG 24-hour run during POWER OPERATION has only a 
small quantitative impact on plant risk. Therefore, the proposed 
change to permit the 24-hour functional test of the EDGs to be 
performed during POWER OPERATION does not significantly increase the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change does not include any physical changes to 
plant design or a change to current Surveillance Requirement 
acceptance criteria. Performance of the Surveillance Requirement 
during POWER OPERATION results in equipment out of service, 
inoperable EDG, which is addressed by current Technical 
Specification limiting condition for operation. Therefore, 
performance of the EDG 24-hour functional testing during POWER 
OPERATION does not create the possibility of a new or different kind 
of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed changes are associated with surveillance 
requirements for the EDGs. The proposed changes allow the EDG 24-
hour functional testing to be performed during POWER OPERATION. 
Performing the functional test during POWER OPERATION will not 
impact the plant design bases or safety analyses because the 
affected EDG will be declared inoperable during the test. During the 
time that the EDG in test is declared inoperable, the system is 
considered to be exempt from the single failure criterion such that 
adequate emergency power will remain available to support the system 
design bases.
    From a design basis perspective, the inoperable EDG effectively 
represents a single failure for the system. Since the emergency 
power system is designed to accomplish its system safety functions 
with only two of the three EDGs in service, and recovery of a failed 
component is not credited in the plant safety analysis (i.e., the 
single failure remains in effect for the entire accident sequence), 
removing an EDG from service to perform a 24-hour functional test 
during POWER OPERATION will not reduce the margin of safety assumed 
in the plant safety analyses.
    The Probabilistic Safety Assessment (PSA) results fall below the 
Acceptance Guidelines for TS changes contained in Regulatory Guides 
1.174 and 1.177. Therefore, the risk of performing the EDG 24-hour 
run during POWER OPERATION has only a small quantitative impact on 
plant risk.
    An integrated assessment of the risk impact of performing the 
24-hour functional test during POWER OPERATION for a single 
inoperable EDG has determined that the risk contribution is small 
and is within regulatory guidelines. Therefore, facility operation 
in accordance with the proposed amendments would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit No. 2, Oswego County, New York

    Date of application for amendment: December 26, 2001.
    Brief description of amendment: The licensee proposed to revise 
Table 3.6.1.3-1, ``Secondary Containment Bypass Leakage Paths Leakage 
Rate Limits,'' of the Technical Specifications to re-designate two 
feedwater system air-operated primary containment isolation valves 
(PCIVs) as simple check valves. Upon approval by the NRC staff, the 
licensee would modify the air-operated PCIVs to become simple check 
valves. The simple check valves will perform the same function as the 
air-operated valves during normal and accident conditions. This design 
change only affects the nonsafety-related remote testing and position 
indication design features of the feedwater check valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff reviewed the licensee's analysis and has performed its own, 
which is presented below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment does not affect the probability of 
previously evaluated accidents because the affected PCIVs were not 
presumed to be initiators or precursors of any accident. The modified 
valves will continue to perform the same function as before. The 
modified valves will not alter or prevent the ability of existing 
structures, systems, or components to perform their intended safety or 
accident-mitigating functions depicted in the Updated Final Safety 
Analysis Report. The proposed amendment and the underlying design 
change will not prevent the unit to continue to comply with applicable 
regulatory requirements. As a result, the proposed amendment will not 
alter the conditions or assumptions used in previously evaluated 
accidents, specifically, the feedwater line break accident outside 
containment, and the loss-of-coolant accident.
    Therefore, operation in accordance with the proposed amendment will 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendment would lead to modification of air-
operated PCIVs to become simple check valves. The modified valves will 
continue to perform the same function (i.e., prevent back flow in the 
feedwater line). Furthermore, the modified valves would not alter or 
prevent the ability of structures, systems, or components to perform 
their intended safety or accident mitigating functions. Thus, 
previously evaluated accident scenarios would not be altered by the 
proposed amendment.
    Accordingly, the proposed amendment and the resulting design 
modification do not create any new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    No. The proposed amendment does not change any analysis 
methodology, safety limits or acceptance criteria. The modified valves 
will have the same level of performance as before.
    Therefore, operation in accordance with the proposed amendment will 
not involve a significant reduction in a margin of safety.
    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment 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).

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

    Date of amendment request: December 19, 2001.

[[Page 5330]]

    Description of amendment request: The proposed amendment would 
extend the completion time under Technical Specification (TS) Section 
3.8.4.A to allow replacement of 125 VDC Batteries 1D1 and 1D2 while at 
power (Mode 1). The proposed amendment would add required actions 
3.8.4.A.2.1 and 3.8.4.A.2.2 as one-time-only alternates and a 
conditional note following 3.8.4.A.1 to allow replacement of the 125 
VDC batteries during a 10-day period for each battery. This TS change 
would be applicable one-time only, for each battery.
    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.
    During the replacement of the existing station batteries, a 
temporary battery will provide the same function as the battery 
being removed. Even though this temporary battery will not meet 
seismic requirements, it will be assembled from safety-related Class 
1E cells. The temporary battery will be subjected to surveillance 
testing prior to being utilized to confirm serviceability. The 
respective DC bus will be continuously energized by the existing 
battery charger. A backup swing charger will also be available which 
is a normal part of system configuration.
    This one-time change also requires that required features be 
declared inoperable when the associated 125 VDC source is inoperable 
and the redundant required feature(s) are also inoperable for at 
least four hours. This action is intended to provide assurance that 
a loss of onsite power, during the period that a 125 VDC source is 
inoperable, does not result in a complete loss of safety function of 
critical systems. The completion time is intended to allow the 
operator time to evaluate and repair any discovered inoperabilities.
    Due to the limited duration of the activity, the very low 
probability of a seismic event over this limited extended completion 
time, and the planned implementing contingency actions, a 
significant increase in the probability of an accident previously 
evaluated does not occur. The proposed change does not affect 
accident initiators or precursors, or design assumptions for the 
systems or components used to mitigate the consequences of an 
accident as analyzed in Chapter 15 of the DAEC UFSAR. The other 
division of DC power will remain operable to support design 
mitigation capability. Therefore, the proposed one-time completion 
time TS amendment does 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.
    During the replacement of the existing station batteries, a 
temporary battery will provide the same function as the batteries 
being removed. Even though this temporary battery does not meet the 
seismic requirements, it possesses adequate capacity to fulfill the 
safety-related requirements of supplying necessary power to the 
associated 125VDC bus. Because the temporary battery will perform 
like the station battery that is currently installed, no new 
electrical or functional failure modes are created. The temporary 
battery will be located in the turbine building which is non-
seismic. The temporary battery will not be placed into seismically 
mounted racks. Thus, a seismic failure of this temporary battery is 
possible. The failure, if it does occur, would not create a new or 
different kind of accident from accidents previously evaluated.
    This one-time change also requires that required features be 
declared inoperable when the associated 125 VDC source is inoperable 
and the redundant required feature(s) are also inoperable for at 
least four hours. This action is intended to provide assurance that 
a loss of onsite power, during the period that a 125 VDC source is 
inoperable, does not result in a complete loss of safety function of 
critical systems. The completion time is intended to allow the 
operator time to evaluate and repair any discovered inoperabilities.
    The proposed one-time change does not introduce any new accident 
initiators or precursors or any new design assumptions for those 
systems or components used to mitigate the consequences of an 
accident. Therefore, the possibility of a new or different kind of 
accident from any previously evaluated has not been created. Thus, 
the proposed one-time completion time extension TS amendment does 
not create the possibility of a new of different kind of accident 
from any previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    During the replacement of the existing station batteries, a 
temporary safety-related battery will perform the same function as 
the battery being removed. Even though this battery will not be 
seismically mounted, it will be assembled from safety-related Class 
1E cells. The battery is functionally similar to the safety-related 
battery that is already installed. It will possess adequate capacity 
to fulfill the requirements of the associated 125VDC bus. The 
proposed replacement activity will not prevent the plant from 
mitigating a Design Basis Accident (DBA) during events that result 
in the loss of power from the temporary battery. In these cases, the 
remaining DC power supporting the design mitigation capability will 
be maintained. Due to the limited duration of the activity, the very 
low probability of a seismic event over this limited extended 
completion time, and the planned implementing contingency actions, a 
significant reduction in the margin of safety will not result. The 
associated DC bus will always be supplied by either the temporary 
battery and/or the battery charger at all times. In addition a spare 
swing battery charger is available. As a result, there is no 
significant reduction in the margin of safety.
    This one-time change also requires that required features be 
declared inoperable when the associated 125 VDC source is inoperable 
and the redundant required feature(s) are inoperable for at least 
four hours. This action is intended to provide assurance that a loss 
of onsite power, during the period that a 125 VDC source is 
inoperable, does not result in a complete loss of safety function of 
critical systems. The completion time is intended to allow the 
operator time to evaluate and repair any discovered inoperabilities.
    Therefore, this proposed amendment does not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: July 30, 2001, as supplemented September 
7, October 16, and December 5, 2001, and January 18, 2002.
    Description of amendment request: This notice supercedes a notice 
published on November 14, 2001 (66 FR 57123).
    The proposed amendments would revise Technical Specification 
5.5.12, ``Primary Containment Leakage Rate Testing Program,'' to allow 
a one-time deferral of the Type A containment integrated leakage rate 
test (ILRT) at the Susquehanna Steam Electric Station (SSES), Units 1 
and 2. The Unit 1 test would be deferred to no later than May 3, 2007, 
and the Unit 2 test would be deferred to no later than October 30, 
2007, resulting in an extended interval of 15 years for performance of 
the next ILRT at each unit. Additionally the proposed amendments would 
allow a one-time deferral of the drywell-to-suppression chamber bypass 
leakage test, Surveillance Requirement (SR) 3.6.1.1.2, so that it would 
continue to be conducted along with the ILRT, consistent with current 
practice.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 5331]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    The frequency of Type A testing does not change the probability 
of an event that results in core damage or vessel failure. Primary 
containment is the engineered feature that contains the energy and 
fission products from evaluated events. The SSES IPE [Individual 
Plant Examination] documents events that lead to containment 
failure. The frequency of events that lead to containment failure 
does not change because it is not a function of the Type A test 
interval. Containment failure is a function of loss of safety 
systems that shutdown the reactor, provide adequate core cooling, 
provide decay heat removal, and loss of drywell sprays.
    Similarly, the frequency of the SR 3.6.1.1.2 bypass test does 
not change the probability of an event that results in core damage 
or vessel failure since they are not a function of the bypass test.
    The consequences of the evaluated accidents are the amount of 
radioactivity that is released to secondary containment and 
subsequently to the public. Normally, extending a test interval 
increases the probability that a Structure, System, or Component 
will fail. However, NUREG-1493, Performance-Based Containment Leak-
Test Program, states that calculated risks in BWR's is very 
insensitive to the assumed leakage rates. The remaining testing and 
inspection programs provide the same coverage as these tests, and 
will maintain containment leakage at appropriately low levels. Any 
leakage problems will be identified and repairs will be made. 
Additionally, the containment is continuously monitored during power 
operation. Anomalies are investigated and resolved. Thus there is a 
high confidence that [containment] integrity will be maintained 
independent of the Type A test and SR 3.6.1.1.2 bypass test 
frequency.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously analyzed?
    Primary containment is designed to contain energy and fission 
products during and after an event. The SSES IPE identifies events 
that lead to containment failure. The proposed revision to the Type 
A and SR 3.6.1.1.2 test interval does not change this list of 
events. There are no physical changes being made to the plant and 
there are no changes to the operation of the plant that could 
introduce a new failure mode creating an accident or affecting 
mitigation of an accident.
    Therefore, this proposed amendment does not involve a 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed one time extension to the Type A test frequency and 
the frequency of SR 3.6.1.1.2 from 10 to 15 years does not involve a 
significant reduction in margin of safety.
    The tests are performed to ensure the degree of reactor 
containment structural integrity and leak-tightness considered in 
the plant safety analysis is maintained. These proposed changes do 
not affect the degree of leak-tightness nor structural integrity of 
the containment. These proposed changes only affect the frequency by 
which the tests are performed. The test acceptance criteria are not 
affected.
    The proposed TS changes do not involve a change in the manner in 
which any plant system is operated or controlled.
    The proposed TS changes do not affect the availability of 
equipment associated with containment integrity that is assumed to 
operate in the plant safety analysis.
    The NUREG-1493 generic study of the effects of extending 
containment leakage testing found that a 20-year interval in Type A 
leakage testing resulted in an imperceptible increase in risk to the 
public. PPL analyses determined the total integrated risk and [Large 
Early Release Frequency] LERF increase is not significant. NUREG-
1493 found that, generically, the design containment leakage rate 
contributes a very small amount of individual risk and would have 
minimal affect since most potential leakage paths are detected by 
Type B and Type C testing. Type B and Type C testing combined with 
visual inspection programs will maintain containment leakage at 
appropriately low levels.
    The vacuum breaker leakage test (SR 3.6.1.1.3) and stringent 
acceptance criteria, combined with the negligible non-vacuum breaker 
leakage area and thorough periodic visual inspection, provide an 
equivalent level of assurance as the SR 3.6.1.1.2 bypass test. PPL 
analyses determined the total integrated risk and LERF increase is 
not significant.
    The combination of the factors described above ensures that the 
proposed changes do not represent a significant reduction on margin 
of safety.

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

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

    Date of amendment request: November 1, 2001.
    Description of amendment request: The proposed changes would modify 
the provisions under which equipment may be considered operable when 
either its normal or emergency power source is inoperable. Technical 
Specifications (TSs) Section 3.0.5 will be deleted under this proposal, 
and additional limiting conditions for operation (LCO) will be 
incorporated into electrical power systems TS 3.8.1.1, A.C. Sources--
Operating. The corresponding TS Bases will be modified accordingly. The 
proposed changes are consistent with the recommendations contained in 
NUREG-1431, Rev. 2, ``Standard Technical Specifications for 
Westinghouse Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The design of the AC electrical power system ensures that 
sufficient power will be available for engineered safeguards 
equipment required for safe shutdown of the facility and mitigation 
of accident conditions. Initial conditions of design basis accidents 
and transients in the Accident Analysis assume required engineered 
safeguards systems are operable and will function in order to 
maintain plant response within design limits. The proposed changes 
to action times do not affect the probability that any accident will 
occur. Since the minimum configuration of equipment assumed in the 
Accident Analysis will remain available, there will similarly be no 
increase in consequences of any accident.
    The proposed changes to action times are consistent with the 
Westinghouse Standard Technical Specification (STS) requirements. 
This specification is intended to provide assurance that an event 
coincident with a failure of the associated normal or emergency 
power supply will not result in complete loss of safety function of 
critical required systems. The completion time allows the operator 
time to evaluate and repair any discovered inoperability. The given 
time periods are considered acceptable because they minimize risk 
while allowing time for restoration before subjecting the unit to 
transients associated with shutdown. These completion times take 
into account the capacity and capability of the remaining AC 
sources, a reasonable time for repairs and the low probability of a 
design basis accident occurring during this period.
    With failure of one offsite power source, the remaining operable 
offsite circuit and diesel generators (DG) are adequate to supply 
electrical power to the onsite Class 1E electrical distribution 
system. At least one complete train of equipment will continue to 
operate in the same manner as assumed in the analyses to mitigate a 
design basis

[[Page 5332]]

accident, given a failure of one component in a redundant train.
    With both required offsite circuits inoperable, onsite emergency 
AC sources remain available to maintain the unit in a safe shutdown 
condition in the event of a design basis accident (DBA) or 
transient. The action completion time is reduced to 12 hours in this 
case. At least one complete train of equipment will operate as 
assumed in the analyses to mitigate a design basis accident, given a 
failure of one component in a redundant train.
    With a single emergency diesel generator inoperable, the 
remaining operable DG and offsite power circuits are adequate to 
supply power to the onsite Class 1E electrical distribution system. 
Required actions ensure that a loss of offsite power during this 
period does not result in a complete loss of safety functions. Four 
hours is considered an acceptable time period to minimize risk 
during this condition, while allowing reasonable time for repair.
    In any of these scenarios at least one train of equipment will 
be available to mitigate an accident and bring the plant to a safe 
shutdown condition, as assumed in the Accident Analysis. There will 
be no impact to radiological dose consequences.
    Therefore, there will be no significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    Expanding the allowable out of service time consistent with 
requirements of Standard Technical Specifications does not introduce 
any new or different failure from any previously evaluated or change 
the manner in which safety systems are operated. The associated 
system and equipment configurations are no different from those 
previously evaluated. The change in allowable action times have been 
considered and determined to be acceptable, without causing 
additional risk. The conditions of TS 3.8.1 continue to ensure that 
an event coincident with a failure of the associated normal or 
emergency power supply will not result in complete loss of safety 
function of critical required systems.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    3. Involve a significant reduction in a margin of safety.
    The power sources and distribution systems are designed to 
ensure sufficient power is available to supply safety related 
equipment required for safe shutdown of the facility and mitigation 
and control of accident conditions. Operability requirements are 
consistent with initial conditions assumed in the accident analysis. 
The proposed changes continue to provide assurance that an event 
coincident with failure of an associated diesel generator or offsite 
power circuit will not result in complete loss of safety function of 
critical required redundant systems or equipment.
    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395 Virgil C. Summer Nuclear Station, 
Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: August 7, 2001.
    Description of amendment request: SCE&G proposes a change to Table 
3.3-2 of the Virgil C. Summer Nuclear Station Technical Specifications 
Surveillance Requirements to include a response time requirement of 0.5 
seconds for the Source Range (SR) Neutron Flux Reactor Trip. The 
proposed change results from SCE&G's review of Westinghouse Nuclear 
Safety Advisory Letter NSAL-00-016. This NSAL notified SCE&G that the 
SR Neutron Flux Reactor Trip is implicitly credited within the accident 
analyses for the Uncontrolled Rod Cluster Control Assembly Bank 
Withdrawal from Subcritical event during Modes 3, 4, and 5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    This change enhances the operability requirements of the SR 
Neutron Flux Instrumentation (NI) system by requiring response time 
testing. The performance of the required response time testing for 
the SR Neutron Flux Channels does not contribute to the initiation 
of any accident previously evaluated. Testing will be done during 
normal channel calibration when the SR Reactor Trip function is not 
required to be operable. During and following the required response 
time testing, there will be no adverse affect on the design and 
operation of the NSSS, BOP, and fluid and auxiliary system which are 
important to safety. Since the reactor coolant pressure boundary 
integrity and normally operating systems are not adversely impacted, 
the probability of occurrence of an accident evaluated in the VCSNS 
FSAR is no greater than the original design basis of the plant.
    The availability of a reactor trip on the SR trip function with 
a defined response time of 0.5 seconds ensures that the event 
consequences of a RWFS event in Modes 3, 4, or 5 remain bounded by 
the current FSAR analysis. This is accomplished by ensuring that the 
reactor is shutdown before any significant power is generated.
    With this change, periodic time response testing of the SR 
reactor trip function will be required to demonstrate that SR 
reactor trip function can be completed within the time limit assumed 
in the accident analyses. This enhanced operability requirement of 
the SR NI system provides additional assurance that the plant will 
be operated within its design and licensing basis. Any event that 
requires the mitigative function of this system will remain bounded 
by the analysis documented in Chapter 15 of the FSAR. No adverse 
hardware, software, setpoint or procedure changes are associated 
with this change. Furthermore, during and following the required 
response time testing, there will be no adverse affect on the design 
and operation of the NSSS, BOP, and fluid and auxiliary systems 
which are important to safety. Given the above, there is no 
potential for additional releases as a result of this activity. 
Therefore, no increase in any previously evaluated accident 
consequences will occur.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Enhancing the operability requirement for a Reactor Protection 
System input can not be considered an accident precursor. This 
change adds response time testing to the SR NI system which assures 
that the accident analysis, including assumptions, is maintained. No 
hardware, software, operational practices or instrumentation 
setpoints are being revised. No change to plant operating 
characteristics or philosophy result from this change. Therefore, 
the possibility of an accident of a different type is not being 
created.
    3. Does this change involve a significant reduction in margin of 
safety?
    TS Table 3.3-2 currently states that the response time for the 
SR NI is not applicable. However, the inherent assumption that this 
system will be the principal system to mitigate the rod withdrawal 
from subcritical accident is described in FSAR 15.2.1. The margin of 
safety is enhanced by the addition of an administrative requirement, 
to assure the safety analysis assumptions are satisfied. The maximum 
response time of 0.5 seconds is consistent with the maximum for 
Power Range and is conservative enough to limit the potential 
excursion to a safe value prior to tripping the plant.

    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: Thomas G. Eppink, South Carolina Electric & 
Gas

[[Page 5333]]

Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard Laufer, Acting.

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

    Date of amendment requests: January 9, 2002.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification 5.4, Technical Specifications (TS) Bases 
control. Specifically, TS 5.4.2 and TS 5.4.2.b would be revised to 
replace the word ``involve'' with ``require'' and delete the term 
``unreviewed safety question,'' respectively. The proposed changes are 
pursuant to the revised regulations in Title 10, Code of Federal 
Regulations (10 CFR) Section 50.59 which eliminated the term 
``unreviewed safety question.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces the word ``involve'' with 
``require'' and deletes reference to the term ``unreviewed safety 
question'' consistent with 10 CFR 50.59. Deletion of the term 
``unreviewed safety question'' was approved by the Nuclear 
Regulatory Commission (NRC) with the revision to 10 CFR 50.59. 
Consequently, the probability of an accident previously evaluated is 
not significantly increased. Changes to the Technical Specification 
(TS) Bases are still evaluated in accordance with 10 CFR 50.59. As a 
result, the consequences of any accident previously evaluated are 
not significantly affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing plant operation. These changes are 
considered administrative changes and do not modify, add, delete, or 
relocate any technical requirements in the TS.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes will not reduce a margin of safety because 
they have no effect on any safety analyses assumptions. Changes to 
the TS Bases that result in meeting the criteria in paragraph (c)(2) 
of 10 CFR 50.59 will still require NRC approval. The proposed 
changes to TS 5.4.2 are considered administrative in nature based on 
the revision to 10 CFR 50.59.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: January 4, 2002.
    Description of amendment request: The proposed amendment would 
change the Safety Limit Minimum Critical Power Ratio (SLMCPR) for 
single loop operation (SLO) in Technical Specification (TS) 2.1.1.2 to 
reflect the results of a cycle-specific calculation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specification [TS] change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The derivation of the revised SLO [single loop operation] SLPCPR 
for [safety limit critical power ratio] Plant Hatch Unit 1 Cycle 21 
for incorporation into the TS, and its use to determine cycle-
specific thermal limits, has been performed using NRC-approved 
methods and procedures. The procedures incorporate cycle-specific 
parameters and reduced power distribution uncertainties in the 
determination of the value for the SLMCPR. These calculations do not 
change the method of operating the plant and have no effect on the 
probability of an accident initiating event or transient.
    The basis of the MCPR Safety Limit is to ensure no mechanistic 
fuel damage is calculated to occur if the limit is not violated. The 
new SLO SLMCPR preserves the existing margin to transition boiling 
and the probability of fuel damage is not increased. Therefore, the 
proposed change does not involve an 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 proposed change is the result of a cycle-specific 
application of NRC-approved methods to the Unit 1 Cycle 21 core 
reload. This change does not involve any new method for operating 
the facility and does not involve any facility modifications. No new 
initiating events or transients result from this change. Therefore, 
the proposed TS change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. Cycle-specific SLMCPRs are calculated using NRC-approved 
methods and procedures, and meet the current fuel design and 
licensing criteria. The SLO SLMCPR will be high enough to ensure 
that greater than 99.9% of all fuel rods in the core are expected to 
avoid transition boiling if the limit is not violated, thereby 
preserving the fuel cladding integrity. Therefore, the proposed TS 
change does not involve a reduction in the margin of safety.

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

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

    Date of amendment request: 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

[[Page 5334]]

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 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    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: October 24, 2001.
    Description of amendment request: The proposed amendment would 
relocate various Technical Specifications (TSs) to the Technical 
Requirements Manual (TRM). Their associated Bases will also be 
relocated to the TRM to be consistent with relocation of the various 
TSs. In addition, the proposed amendment corrects various typographical 
and page numbering errors, deletes an outdated one-time exception, and 
makes minor formal changes to improve consistency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This request involves relocation of information to the Technical 
Requirements Manual and administrative changes only. No actual plant 
equipment or accident analyses will be affected by the proposed 
changes. Therefore, the proposed amendment does not result in any 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    This request involves relocation of information to the Technical 
Requirements Manual and administrative changes only. The proposed 
change does not alter the performance of the equipment or the manner 
in which the equipment will be operated. The equipment will still be 
verified by test, if applicable, in accordance with applicable 
surveillance requirements. Changing the location of these 
requirements and surveillances from Technical Specifications to the 
Technical Requirements Manual will not create any new accident 
initiators or scenarios. Since the proposed changes only allow 
activities that are presently approved and conducted, no possibility 
exists for a new or different kind of accident from those previously 
evaluated.
    3. Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    This request involves relocation of information to the Technical 
Requirements Manual and administrative changes only. No actual plant 
equipment or accident analyses will be affected by the proposed 
change. Additionally, the proposed changes will not relax any 
criteria used to establish safety limits, will not relax any safety 
systems settings, or will not relax the bases for any limiting 
conditions of operation. Therefore, the proposed changes will not 
impact the margin of safety.

    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 &

[[Page 5335]]

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: November 5, 2001.
    Description of amendment request: The proposed amendments would 
revise specific requirements of Technical Specification (TS) Section 
6.0, ``Administrative Controls.'' The proposed amendments include 
relocating specific TS administrative control requirements to licensee-
controlled documents; updating specific management titles to more 
generic title positions; updating requirements to be consistent with 
current industry standards; and reformatting, renumbering, and 
rewording existing requirements for better readability. The proposed 
changes include Items 1 thru 125, and 127 in Table 1 of Attachment 1 of 
the licensee's submittal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve reformatting, renumbering, and 
rewording of the existing TS. These modifications involve no 
technical changes to the existing TS. As such, these changes are 
administrative in nature and do not effect initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes involve reformatting, renumbering, and 
rewording of the existing TS. The changes do not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or changes in methods governing normal plant 
operation. The changes will not impose any new or different 
requirements or eliminate any existing requirements. Therefore, the 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed changes involve reformatting, renumbering, and 
rewording of the existing TS. The changes are administrative in 
nature and will not involve any technical changes. The changes will 
not reduce a margin of safety because they have no impact on any 
safety analysis assumptions. Also, since these changes are 
administrative in nature, no question of safety is involved. 
Therefore, the changes do not involve a significant reduction in a 
margin of safety.

More Restrictive Changes

    The proposed changes designated as ``More Restrictive'' (M) 
technical changes involve adding more restrictive requirements to 
the existing TS by either making current requirements more stringent 
or by adding new requirements that currently do not exist. These 
changes have been evaluated to not be detrimental to plant safety. 
These changes are modifications of requirements to provide 
consistency with the Improved Standard Technical Specifications 
recommended in NUREG-1431. The proposed changes include Items 39, 
51, 129 and 130 in Table 1.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes provide more stringent requirements for 
operation of the facility. The more stringent requirements do not 
result in operation that will increase the probability of initiating 
an analyzed event and do not alter assumptions relative to 
mitigation of an accident or transient event. The more stringent 
requirements continue to ensure process variables, structures, 
systems, and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The changes 
will not impose any new or different requirements or eliminate any 
existing requirements. Therefore, the changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The imposition of more stringent requirements either has no 
impact on or increases the margin of plant safety. As noted in the 
discussion of the changes, each change in this category, by 
definition, provides additional restrictions to enhance plant 
safety. The changes maintain requirements within the safety analyses 
and licensing basis. Therefore, these changes do not involve a 
significant reduction in a margin of safety.

Less Restrictive Changes L.1

    Current TS 6.8.3.i, ``Diesel Fuel Oil Testing Program,'' 
requires properties for ASTM 2D fuel oil to be within limits within 
30 days following sampling. The proposed change will increase the 
time in which compliance must be verified following sampling from 30 
days to 31 days. This change is reasonable based on the relatively 
small increase in time and the probability of a major problem being 
found that would prevent the diesel generator from starting and 
operating. The proposed change, Item 70 in Table 1, is consistent 
with NUREG-1431.
    In accordance with the criteria set forth in 10 CFR 50.92, the 
South Texas Project has evaluated this proposed TS change and 
determined that it involves no significant hazards consideration. 
The following is provided in support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change extends the allowed completion time from 30 
days to 31 days to verify that diesel fuel sample properties comply 
with ASTM 2D. This change does not affect the probability of an 
accident. Diesel fuel oil is not an initiator of any analyzed event. 
The consequences of an accident are not increased significantly 
because of the remote probability of an event occurring during the 
24-hour period. Also, the probability of a major problem being found 
which would prevent the diesel generator from starting and operating 
is remote. The change will not alter the ability to mitigate an 
accident or transient event. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The change 
will not impose any new or different requirements or eliminate any 
existing requirements. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change extends the allowed completion time from 30 
days to 31 days to verify that diesel fuel sample properties comply 
with ASTM 2D. The change does not significantly decrease the margin 
of safety because of the remote probability of an event occurring 
during the 24-hour period. Also, the probability of a major problem 
being found which would prevent the diesel generator from starting 
and operating is remote. The safety analysis assumptions will still 
be maintained. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

Less Restrictive Change L.2

    Current TS 6.9.1.2 and 6.9.1.2.a require annual submittal of an 
Occupational Radiation Exposure Report by March 1 of the calendar 
year following the exposures. The submittal date is revised to April 
30. This change is consistent with previous comprehensive revisions 
to 10 CFR Part 20. The report is provided to supplement the 
information required by 10 CFR 20.2206(b), which is filed on or 
before April 30 in accordance with 10 CFR 20.2206(c). The proposed 
change, Item 76 in Table 1, is consistent with NUREG-1431.

[[Page 5336]]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not result in any changes in hardware 
or methods of operation. The change in date for submittal of ``after 
the fact'' information is not considered in the safety analysis and 
cannot initiate or affect the mitigation of an accident in any way. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The change 
will impact only the administrative requirements for submittal of 
information and does not directly impact the operation of the plant. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not impact the margin of safety since 
the margin of safety is not dependent on the submittal of 
information. The safety analysis assumptions will still be 
maintained. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

Less Restrictive Change L.3

    Current TS 6.9.1.3 requires annual submittal of a Radiological 
Environmental Operating Report by May 1 of each year. The submittal 
date is revised to May 15. This is an interval increase of 15 days. 
There is no requirement for the NRC to approve this report and 10 
CFR [Part] 50 does not specify a specific reporting date. The 
proposed change, Item 82 in Table 1, is consistent with NUREG-1431.
    In accordance with the criteria set forth in 10 CFR 50.92, the 
South Texas Project has evaluated this proposed TS change and 
determined that it involves no significant hazards consideration. 
The following is provided in support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not result in any changes in hardware 
or methods of operation. The change in date for submittal of ``after 
the fact'' information is not considered in the safety analysis and 
cannot initiate or affect the mitigation of an accident in any way. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The change 
will impact only the administrative requirements for submittal of 
information and does not directly impact the operation of the plant. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not impact the margin of safety since 
the margin of safety is not dependent on the submittal of 
information. The safety analysis assumptions will still be 
maintained. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

Less Restrictive Change L.4

    Current TS 6.9.1.4 requires annual submittal of a Radioactive 
Effluent Release Report within 60 days after January 1 of each year. 
The submittal date is revised to May 1. This is an interval increase 
of approximately 60 days. The proposed change, Item 85 in Table 1, 
is consistent with NUREG-1431.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not result in any changes in hardware 
or methods of operation. The change in date for submittal of ``after 
the fact'' information is not considered in the safety analysis and 
cannot initiate or affect the mitigation of an accident in any way. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The change 
will impact only the administrative requirements for submittal of 
information and does not directly impact the operation of the plant. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not impact the margin of safety since 
the margin of safety is not dependent on the submittal of 
information. The safety analysis assumptions will still be 
maintained. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
Less Restrictive Change L.5

    The details specifying responsibility for initiating the 
Radiation Work Permit (RWP) surveillance frequency are being 
deleted. The requirement of current TS 6.12.1.c pertains to the 
individual qualified in radiation protection responsible for 
providing control over the activities in a high radiation area, 
including the performance of periodic radiation surveillances. The 
details specifying responsibility for the surveillance frequency in 
the RWP have no bearing on the requirements for entering a high 
radiation area. RWP details are controlled by plant procedures. 
Deleting these details eliminates ambiguity in the TS and the 
possibility for a misinterpretation of the TS requirements. The 
proposed change is provided in Table 1 as Item 103.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change eliminates ambiguity in the TS details 
specifying responsibility for the surveillance frequency in the 
Radiation Work Permit. The proposed change does not result in any 
changes in hardware or methods of operation. The details pertaining 
to the surveillance frequency in the Radiation Work Permit are not 
considered in the safety analysis and cannot initiate or affect the 
mitigation of an accident in any way. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed [change] does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The change 
will not impose any new or different requirements or eliminate any 
existing requirements. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not impact the margin of safety since 
the margin of safety is not dependent on who initiates the 
surveillance frequency of the Radiation Work Permit. The safety 
analysis assumptions will still be maintained. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

Less Restrictive Change L.6

    The details specifying the individuals responsible for 
performance of the review of the use of overtime are being deleted, 
and the frequency at which the overtime review is performed is being 
changed from monthly to periodic. The details specifying 
responsibility for performance of the overtime review and the 
frequency of review are controlled by plant procedures. The proposed 
changes are consistent with the programmatic controls required by 
NUREG-1431. The proposed changes are provided in Table 1 as Item 
30a.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes delete the details specifying the 
individuals responsible for performance of the overtime use review, 
and changes the frequency at which the overtime review is performed 
from monthly to periodic. The proposed change does not result in any 
changes in hardware or methods of operation. The details pertaining 
to the review of overtime are not considered in the safety analysis 
and cannot initiate or affect the mitigation of an accident in any 
way. Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 5337]]

    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed does not involve a physical alteration of the plant 
(no new or different type of equipment will be installed) or changes 
in methods governing normal plant operation. The change will not 
impose any new or different requirements or eliminate any existing 
requirements. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not impact the margin of safety since 
the margin of safety is not dependent on who performs the overtime 
review, nor on the frequency at which the review is performed. The 
safety analysis assumptions will still be maintained. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

Less Restrictive Change L.7

    The details specifying the actions to be taken in the event a 
Safety Limit is violated are deleted from the Specifications. The 
details regarding notification and reporting to the Commission are 
unnecessary, since reporting requirements are delineated in 10 CFR 
50.72 and 50.73. The details regarding onsite notification 
requirements and review of the report by PORC [Plant Operations 
Review Committee] and NSRB [Nuclear Safety Review Board] are 
unnecessary, since plant policies and procedures already provide 
guidance on onsite notification and review of reports by these 
committees. Furthermore, these notification and reporting 
requirements are beyond the criteria of 10 CFR 50.36(c)(5) for 
inclusion in the Administrative Controls Section of the TS, and 
programmatic controls regarding actions to be taken for Safety Limit 
violations are not included in NUREG-1431. The proposed changes are 
provided in Table 1 as Item 30a.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes delete the details regarding actions to be 
taken in the event of a Safety Limit violation. The proposed change 
does not result in any changes in hardware or methods of operation. 
The details pertaining to notification and reporting of Safety Limit 
violations are not considered in the safety analysis and cannot 
initiate or affect the mitigation of an accident in any way. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed does not involve a physical alteration of the plant 
(no new or different type of equipment will be installed) or changes 
in methods governing normal plant operation. The change will not 
impose any new or different requirements or eliminate any existing 
requirements. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not impact the margin of safety since 
the margin of safety is not dependent on notification and reporting 
of Safety Limit violations. The safety analysis assumptions will 
still be maintained. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

Relocation of Requirements

    The proposed changes designated as ``Relocated'' (R) technical 
changes involve the relocation of existing TS requirements or 
details to other licensee-controlled documents such as the UFSAR 
[Updated Final Safety Analysis Report], TRM [Technical Requirements 
Manual], ODCM [Offsite Dose Calculation Manual], or OQAP 
[Operational Quality Assurance Plan]. Future modification of 
relocated Administrative Controls requirements is adequately 
controlled by regulatory requirements such as 10 CFR 50.59 and 10 
CFR 50.54. The proposed changes include Items 4, 12, 13, 15, 22, 25, 
29, 31, 32, 40, 41, 42, 44, 46, 49, 52, 55, 58, 59, 68, 75, 96, 112, 
117, 118, and 126 in Table 1.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes relocate certain details from the TS to the 
UFSAR, TRM, OQAP, or other licensee-controlled documents. These 
licensee-controlled documents containing the relocated information 
will be maintained in accordance with 10 CFR 50.59 or 10 CFR 50.54, 
as appropriate. The UFSAR is subject to the change control 
provisions of 10 CFR 50.71(e) and the plant procedures and other 
licensee-controlled documents are subject to controls imposed by 
plant administrative procedures, which endorse applicable 
regulations and standards. Since any changes to the UFSAR, TRM, 
OQAP, or other licensee-controlled documents will be evaluated per 
10 CFR 50.59 or 10 CFR 50.54, such changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Therefore, the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve physical alteration of the 
plant (no new or different type of equipment will be installed) or 
change in the methods governing normal plant operation. The proposed 
changes will not impose or eliminate any requirements and adequate 
control of the information will be maintained. Thus, these changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes will not reduce a margin of safety because 
they have no impact on any safety analysis assumptions. In addition, 
the details to be relocated from the TS to the UFSAR, TRM, OQAP, or 
other licensee-controlled documents are the same as in the existing 
TS. Since any future change to these details in the UFSAR, TRM, 
OQAP, or other licensee-controlled documents will be evaluated per 
the requirements of 10 CFR 50.59 or 10 CFR 50.54, as appropriate, 
such changes would not involve a significant reduction in a margin 
of safety. Based on 10 CFR 50.92, the existing requirement for NRC 
review and approval of revisions to these details proposed for 
relocation does not have a specific margin of safety upon which to 
evaluate. However, since the proposed changes are consistent with 
NUREG-1431, which was approved by the NRC Staff, revising the TS to 
reflect the approved level of detail ensures no significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licenses's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    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: December 10, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 4.0.1 and 4.0.3 from the current 
South Texas Project (STP) TS format to the Improved TS format. In 
addition, the licensee has proposed that a Bases Control Program be 
incorporated into Section 6.0 of the TSs in order to (1) specify an 
administrative process for making changes to the TS bases, (2) 
delineate what kinds of changes can be made to the TS Bases without 
prior NRC approval, and (3) to provide for consistency between the TS 
Bases and the STP Final Safety Analysis Report. TS 4.0.3 would also be 
changed to reflect Technical Specification Task Force (TSTF) 358, 
Revision 6, changes to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period would be extended from the current limit of ``* * * up to 
24 hours or up to the limit of the specified surveillance interval, 
whichever is less'' to ``* * * up to 24 hours or up to the limit of the 
specified surveillance interval, whichever is greater.'' The following 
requirement would be added to TS 4.0.3: ``A risk

[[Page 5338]]

evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves rewording of the existing Technical 
Specifications [4.0.1 and 4.0.3] to be consistent with NUREG-1431, 
Revision 2. These modifications involve no technical changes to the 
existing Technical Specifications. As such, these changes are 
administrative in nature and do not affect initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change involves incorporation of the NUREG-1431, 
Revision 2, Bases Control Program requirements into the STP 
Technical Specifications. These modifications involve no technical 
changes to the existing Technical Specifications. As such, these 
changes are administrative in nature and do not affect initiators of 
analyzed events or assumed mitigation of accident or transient 
events. Therefore, these changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change involves rewording of the existing Technical 
Specifications [4.0.1 and 4.0.3] to be consistent with NUREG-1431, 
Revision 2. The change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The changes 
will not impose any new or different requirements or eliminate any 
existing requirements. Therefore, the changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change involves incorporation of the NUREG-1431, 
Revision 2, Bases Control Program requirements into the STP 
Technical Specifications. The changes do not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or changes in methods governing normal plant 
operation. The changes will not impose any new or different 
requirements or eliminate any existing requirements. Therefore, the 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in [a] 
margin of safety?
    The proposed change involves rewording of the existing Technical 
Specifications [4.0.1 and 4.0.3] to be consistent with NUREG-1431, 
Revision 2. The changes are administrative in nature and will not 
involve any technical changes. The changes will not reduce a margin 
of safety because they have no impact on any safety analysis 
assumptions. Also, since these changes are administrative in nature, 
no question of safety is involved. Therefore, the changes do not 
involve a significant reduction in a margin of safety.
    The proposed change involves incorporation of the NUREG-1431, 
Revision 2, Bases Control Program requirements into the STP 
Technical Specifications. The changes are administrative in nature 
and will not involve any technical changes. The changes will not 
reduce a margin of safety because they have no impact on any safety 
analysis assumptions. Also, since these changes are administrative 
in nature, no question of safety is involved. Therefore, the changes 
do not involve a significant reduction in a margin of safety.

    With regard to the changes associated with TSTF-358, Revision 6, 
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 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 10, 2001.
    As required by 10 CFR 50.91(a), an analysis of the issue of no 
significant hazards consideration 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 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.
    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 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.
    3. The proposed change does not involve a significant reduction 
in 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 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.

[[Page 5339]]

    NRC Section Chief: Robert A. Gramm.

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: November 8, 2001 (TS 01-06).
    Brief description of amendments: The proposed amendment would 
revise a License Condition and the Technical Specifications (TS) for 
Sequoyah Units 1 and 2. The proposed change would delete License 
Condition 2.H, ``Reporting to the Commission,'' Administrative Control 
Section 6.6, ``Reportable Event Action,'' and Administrative Control 
Section 6.7, ``Safety Limit Violation.'' Because Administrative Control 
Section 6.6 is referenced in several Limiting Conditions for Operation 
(LCOs) and associated TS Bases, these LCOs and TS Bases would also be 
modified to remove those references.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    These revisions govern the reporting of either site 
characteristics and past events or of events covered under current 
NRC regulations and the proposed amendment is administrative in 
nature. Therefore, it does not increase the probability or 
consequences of any accident previously evaluated because it does 
not affect the state of the plant in any physical manner.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment is strictly administrative and does not 
affect plant equipment or operational procedures. Therefore, it will 
not create any new or different accidents.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment affects the reporting to the Commission. 
As such, it does not affect personnel, public, or plant safety. 
Since the amendment will not affect the plant in a physical manner 
nor will it affect personnel, public, or plant safety, it will 
therefore not reduce the margin of safety.

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

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

    Date of application for amendments: January 15, 2002 (TS 01-13).
    Brief description of amendments: The proposed amendment would 
revise Technical Specification (TS) Section 4.0.5.c to provide an 
exception to the recommendations of Regulatory Position c.4.b of NRC 
Regulatory Guide 1.14, Revision 1, ``Reactor Coolant Pump Flywheel 
Integrity,'' dated August 1975. This change is in accordance with 
Improved Standard TS Generic Change Traveler TSTF-237, Revision 1, 
Westinghouse Electrical Corporation Topical Report WCAP-14535A, 
``Topical Report on Reactor Coolant Pump Flywheel Inspection 
Elimination.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    An integral part of the Reactor Coolant System (RCS) in a 
pressurized water reactor is the RCP [reactor coolant pump]. The RCP 
ensures an adequate cooling flow rate by circulating large volumes 
of the primary coolant water at high temperature and pressure 
through the RCS. Following an assumed loss of power to the RCP 
motor, the flywheel, in conjunction with the impeller and motor 
assembly, provides sufficient rotational inertia to assure adequate 
core cooling flow during RCP coastdown.
    Westinghouse Electric Corporation Topical Report WCAP-14535A, 
``Topical Report on Reactor Coolant Pump Flywheel Inspection 
Elimination,'' dated November 1996, provides the technical basis for 
the elimination of inspection requirements for RCP flywheels for all 
domestic Westinghouse plants. In the Safety Evaluation for WCAP-
14535, dated September 1996, the NRC stated that the evaluation 
methodology described in WCAP-14535 is appropriate and the criteria 
are in accordance with the design criteria of RG 1.14.
    RCP flywheel inspections have been performed for 20 years with 
no indications of service induced flaws. Flywheel integrity 
evaluations show a very high flaw tolerance for the RCP flywheels. 
Crack extension over a 60-year service life is negligible. 
Structural reliability studies have shown that eliminating 
inspections after 10 years of plant life will not significantly 
change the probability of failure.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed change does not 
alter or prevent the ability of structures, systems, and components 
(SSC) from performing their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the SQN Updated Final Safety Analysis Report (UFSAR). 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 SQN UFSAR.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not modify the design or function of 
the RCP flywheels. Based upon the results of WCAP-14535A, no new 
failure mechanisms will be introduced by the revised RCP Flywheel 
Inservice Inspection Program. As presented in WCAP-14535A, detailed 
stress analysis and risk assessments have been performed that 
indicate that there would be no change in the probability of failure 
for RCP flywheels if all inspections were eliminated. In addition, 
the flywheel integrity evaluations show that RCP flywheels exhibit a 
very high tolerance for the presence of flaws.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    There is no significant mechanism for in-service degradation of 
the flywheels since they are isolated from the primary coolant 
environment. Additionally, WCAP-14535A analyses have shown there is 
no significant deformation of the flywheels even at maximum 
overspeed conditions. Likewise, the results of RCP flywheel 
inspections performed throughout the industry and at SQN identified 
no indications that would affect flywheel integrity.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 5340]]

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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 26, 2001.
    Brief description of amendments: The proposed amendments would 
revise Technical Specifications (TS) 5.5.16, ``Containment Leakage Rate 
Testing Program'' to allow for a one-time extension of the current 
interval between the Type A tests from 10 to 15 years.
    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 revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10CFR50, 
Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 10 years, based on past performance, would be 
extended on a one time basis to 15 years from the last Type A test. 
The proposed extension to Type A testing does not involve a 
significant increase in the consequences of an accident since 
research documented in NUREG-1493, ``Performance-Based Containment 
System Leakage Testing Requirements,'' September 1995, has found 
that, generically, very few potential containment leakage paths are 
not identified by Type B and C tests. The NUREG concluded that 
reducing the Type A testing frequency to one per twenty years was 
found to lead to an imperceptible increase in risk. A high degree of 
assurance is provided through testing and inspection that the 
containment will not degrade in a manner detectable only by Type A 
testing. The last Type A test show[s] leakage to be below acceptance 
criteria, indicating a very leak tight containment. Inspections 
required by the American Society of Mechanical Engineers (ASME) Code 
[Boiler and Pressure Vessel Code] Section XI (Subsections IWE and 
IWL) and maintenance rule monitoring (10CFR50.65, ``Requirements for 
Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants'') are performed in order to identify indications of 
containment degradation that could affect that leak tightness. Type 
B and C testing required by Technical Specifications will identify 
any containment opening such as valves that would otherwise be 
detected by the Type A tests. These factors show that a Type A test 
extension will not represent a significant increase in the 
consequences of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10CFR50, 
Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 10 years, based on past performance, would be 
extended on a one time basis to 15 years from the last Type A test. 
The proposed extension to Type A testing cannot create the 
possibility of a new or different type of accident since there are 
no physical changes being made to the plant and there are no changes 
to the operation of the plant that could introduce a new failure 
mode creating an accident or affecting the mitigation of an 
accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10CFR50, 
Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 10 years, based on past performance, would be 
extended on a one time basis to 15 years from the last Type A test. 
The proposed extension to Type A testing will not significantly 
reduce the margin of safety. The NUREG-1493, ``Performance-Based 
Containment System Leakage Testing Requirements,'' September 1995, 
generic study of the effects of extending containment leakage 
testing found that a 20 year extension in Type A leakage testing 
resulted in an imperceptible increase in risk to the public. NUREG-
1493 found that, generically, the design containment leakage rate 
contributes about 0.1 percent to the individual risk and that the 
decrease in Type A testing frequency would have a minimal affect on 
this risk since 95% of the potential leakage paths are detected by 
Type C testing. Regular inspections required by the American Society 
of Mechanical Engineers (ASME) Code Section XI (Subsections IWE and 
IWL) and maintenance rule monitoring (10CFR50.65, ``Requirements for 
Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants'') will further reduce the risk of a containment leakage path 
going undetected.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.

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

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 proposed amendment would 
revise Required Actions for Limiting Conditions for Operation (LCOs) 
3.3.1, ``Reactor Trip (RTS) Instrumentation;'' 3.3.9, ``Boron Dilution 
Mitigation System (BDMS);'' 3.4.5, ``RCS Loops--MODE 3;'' 3.4.6, ``RCS 
Loops--MODE 4;'' 3.4.7, ``RCS Loops--MODE 5, Loops Filled;'' 3.4.8, 
``RCS Loops--MODE 5, Loops Not Filled;'' 3.8.2, ``AC Sources--
Shutdown;'' 3.8.5, ``DC Sources--Shutdown;'' 3.8.8, ``Inverters--
Shutdown;'' 3.8.10, ``Distribution Systems--Shutdown;'' 3.9.3, 
``Nuclear Instrumentation;'' 3.9.5, ``Residual Heat Removal (RHR) and 
Coolant Circulation--High Water Level;'' and 3.9.6, ``Residual Heat 
Removal (RHR) and Coolant Circulation--Low Water Level'' in the 
Callaway Plant Technical Specifications (TSs). The Required Actions 
proposed to be revised require suspension of operations involving 
positive reactivity additions or reactor coolant system (RCS) boron 
concentration reductions. In addition, the proposed amendment would 
revise Notes, for several of the LCOs, that preclude reductions in RCS 
boron concentration. This amendment would revise these Required Actions 
and LCO Notes to allow small, controlled, safe insertions of positive 
reactivity, but limits the introduction of positive reactivity such 
that compliance with the required shutdown margin or refueling boron 
concentration limits will still be satisfied.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. The RTS instrumentation and reactivity control 
systems will be unaffected. Protection systems will continue to 
function in a manner consistent with the plant design basis. All 
design, material, and construction

[[Page 5341]]

standards that were applicable prior to the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the FSAR [Final Safety Analysis Report] are not 
adversely affected because the changes to the Required Actions and 
LCO Notes assure the limits on SDM [shutdown margin] and refueling 
boron concentration continue to be met, consistent with the analysis 
assumptions and initial conditions included within the safety 
analysis and licensing basis. The activities covered by this 
amendment application are routine operating evolutions. The proposed 
changes do not reduce the capability of reborating the RCS.
    The proposed changes will not involve a significant increase in 
the probability of any event initiators. The initiating event for an 
inadvertent boron dilution event, as discussed in FSAR Section 
15.4.6, is a failure in the reactor makeup control system (RMCS) or 
operator error such that inventory makeup with the incorrect boron 
concentration enters the RCS by way of the CVCS [chemical volume and 
control system] mixing tee. Since the RMCS design is unchanged, 
there will be no initiating event frequency increase associated with 
equipment failures. However, there could be an increased exposure 
time per operating cycle to potential operator errors during TS 
Conditions that, heretofore, prohibited positive reactivity 
additions. As such, the RTS Instrumentation, BDMS, and RCS Loops TS 
Bases changes from TSTF-286, Revision 2, have been augmented to 
preclude the introduction of reactor makeup water into the RCS via 
the CVCS mixing tee when one source range neutron flux channel (and, 
thus, the associated BDMS train) is inoperable or when no RCS loop 
is in operation. The equipment and processes used to implement RCS 
boration or dilution evolutions are unchanged and the equipment and 
processes are commonly used throughout the applicable MODES under 
consideration. There will be no degradation in the performance of, 
or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident situation. 
There will be no change to normal plant operating parameters or 
accident mitigation performance. Required Action A.1 of LCO 3.3.9 
limits the exposure to one inoperable BDMS train, which may be 
caused by an inoperable source range neutron flux channel. During 
the time the plant is in a TS Condition with a finite equipment 
restoration time, a single failure of the opposite train is not 
postulated. However, administrative controls have been added to this 
Action's Bases to highlight the need for operator awareness during 
all reactivity manipulations and to preclude introduction of reactor 
makeup water into the RCS.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the FSAR.
    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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation or change any operating limits. The proposed changes 
merely permit the conduct of normal operating evolutions when 
additional controls over core reactivity are imposed by the 
Technical Specifications. The proposed changes do not introduce any 
new equipment into the plant or alter the manner in which existing 
equipment will be operated. The changes to operating procedures are 
minor, with clarifications provided that required limits must 
continue to be met. No performance requirements or response time 
limits will be affected. These changes are consistent with 
assumptions made in the safety analysis and licensing basis 
regarding limits on SDM and refueling boron concentration.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the design or performance of the 
7300 Process Protection System, Nuclear Instrumentation System, or 
Solid State Protection System used in the plant protection systems.
    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 do not alter the limits on SDM or refueling 
boron concentration. The nominal trip setpoints specified in the 
Technical Specifications Bases and the safety analysis limits 
assumed in the transient and accident analyses are unchanged. None 
of the acceptance criteria for any accident analysis is changed.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (FH), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    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.

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

    Date of application request: December 13, 2001.
    Description of amendment request: The amendment would revise the 
Limiting Condition for Operation (LCO) 3.5.5, Required Action A.1 for 
the LCO, and Surveillance Requirement 3.5.5.1 in Technical 
Specification (TS) 3.5.5, ``Seal Injection Flow.'' The revision would 
replace the flow and differential pressure limits for the reactor 
coolant pump (RCP) seal injection flow stated in TS 3.5.5 by limits in 
Figure 3.5.5-1 that would be added to TS 3.5.5.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. The RTS [reactor trip system] instrumentation 
and reactivity control systems will be unaffected. Protection 
systems will continue to function in a manner consistent with the 
plant design basis. All design, material, and construction standards 
that were applicable prior to the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the FSAR [Final Safety Analysis Report] are not 
adversely affected because the changes continue to assure the 
analysis assumptions and initial conditions included within the 
safety analysis and licensing basis are satisfied.
    The proposed changes will not involve a significant increase in 
the probability of any event initiators. The initiating event for a 
loss of coolant accident, as discussed in FSAR Section 15.6.5, is a 
break in the RCS [reactor coolant system] piping. Since the RCS 
piping design is unchanged, there will be no initiating event 
frequency increase associated with pipe breaks. There will be no 
degradation in the performance of, or an increase in the number of 
challenges imposed on, safety-related equipment assumed to function 
during an accident situation. There will be no change to normal 
plant operating parameters or accident mitigation performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the FSAR.

[[Page 5342]]

    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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation. The proposed changes do not introduce any new equipment 
into the plant or alter the manner in which existing equipment will 
be operated. The changes to operating procedures are minor, with 
clarifications provided that required limits must continue to be 
met. No performance requirements or response time limits will be 
affected. These changes are consistent with assumptions made in the 
safety analysis and licensing basis regarding limits on RCP seal 
injection flow.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the design or performance of the 
7300 Process Protection System, Nuclear Instrumentation System, or 
Solid State Protection System used in the plant protection systems. 
Therefore, the proposed changes do 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 changes do not alter the input parameters listed in 
FSAR Table 15.6-9 and used in large break and small break LOCA 
[loss-of-coolant accident] peak cladding temperature analyses. The 
containment pressure and temperature analyses are not adversely 
impacted. The nominal reactor and ESFAS [engineered safety feature 
actuation system] trip setpoints (Technical Specification Bases 
Tables B 3.3.1-1 and B 3.3.2-1), reactor and ESFAS allowable values 
(Technical Specification Tables 3.3.1-1 and 3.3.2-1), and the safety 
analysis limits assumed in the transient and accident analyses (FSAR 
Table 15.0-4) are unchanged. None of the acceptance criteria for any 
accident analysis is changed.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protective functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (FH), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    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.

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

    Date of amendment request: April 3, 2001 as supplemented by letters 
dated October 22 and December 18, 2001. The April 3, 2001, amendment 
application was previously noticed in the Federal Register on May 2, 
2001 (66 FR 22036).
    Description of amendment request: The supplemental letter of 
October 22, 2001, added the following change to the technical 
specifications (TSs): revise TS Section 5.6.5 by adding TS 2.1.1 on 
reactor core safety limits on the existing list of core operating 
limits for each reload cycle that are documented in the Core Operating 
Limits Report (COLR). This proposed change is being added to the 
previous changes requested by the licensee's letter of April 3, 2001. 
The amendment would make the following changes to the TSs:
    (1) Revise Safety Limit 2.1.1 by replacing Figure 2.1.1-1, 
``Reactor Core Safety Limits,'' with a reference to limits being 
specified in the Core Operating Limits Report (COLR) and by adding two 
reactor core safety limits on departure from nucleate boiling ratio 
(DNBR) and peak fuel centerline temperature.
    (2) Revise Note 1 on the over temperature T in Table 
3.3.1-1 of TS 3.3.1, ``Reactor Trip System Instrumentation,'' by 
replacing values of parameters with a reference to the values being 
specified in the COLR and correcting the expression for one term in the 
inequality for over temperature T.
    (3) Revise Note 2 on the overpower T in Table 3.3.1-1 by 
replacing values of parameters with a reference to the values being 
specified in the COLR.
    (4) Replace the limits for the reactor coolant system (RCS) 
pressure and average temperature with a reference to the limits being 
specified in the COLR for Limiting Condition for Operation (LCO) 3.4.1 
and Surveillance Requirements (SRs) 3.4.1.1 and 3.4.1.2.
    (5) Add the phrase ``and greater than or equal to the limit 
specified in the COLR'' to the RCS total flow rate in LCO 3.4.1 and SRs 
3.4.1.3 and 3.4.1.4.
    (6) Move items a. and b. to the left in the Note to the 
applicability in LCO 3.4.1.
    (7) Revise TS Section 5.6.5 by adding TS 2.1.1 on reactor core 
safety limits, TS 3.3.1 on over temperature and overpower T 
trip setpoints, and TS 3.4.1 on RCS pressure, temperature, and flow 
limits to the existing list of core operating limits for each reload 
cycle that are documented in the COLR and revising the list of topical 
reports in the COLR that represent the analytical methods approved by 
the Commission to determine core operating limits.
    The proposed changes remove cycle-specific parameter limits and 
relocate them to the COLR, but they do not change any of the limits. 
The changes add more specific requirements regarding DNBR limit and 
peak fuel centerline temperature limit to the TSs, revise the list of 
topical reports in the list of NRC-approved analytical methods, correct 
one term of an expression, and move terms in a Note to the mode 
applicability for an LCO.
    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 are programmatic and administrative in 
nature which do not physically alter safety related systems, nor 
affect the way in which safety related systems perform their 
functions. More specific requirements regarding the safety limits 
(i.e., DNBR limit and peak fuel centerline temperature limit) are 
being imposed in TS 2.1.1, ``Reactor Core Safety Limits,'' which 
replace the Reactor Core Safety Limits figure and are consistent 
with the values stated in the USAR [Updated Safety Analysis Report]. 
The proposed changes remove the cycle-specific parameter limits from 
TS 3.4.1 and relocate them to the COLR which do not change plant 
design or affect system operating parameters. In addition, the 
minimum limit for RCS total flow rate is being retained in TS 3.4.1 
to assure that a lower flow rate than reviewed by the NRC will not 
be used. The proposed changes do not, by themselves, alter any of 
the parameter limits. The removal of the cycle-specific parameter 
limits from the TS does not eliminate existing requirements to 
comply with the parameter limits. The existing TS Section 5.6.5b, 
COLR Reporting Requirements, continues to ensure that the analytical 
methods used to determine the core operating limits meet NRC 
reviewed and approved methodologies. The existing TS Section 5.6.5c, 
COLR Reporting

[[Page 5343]]

Requirements, continues to ensure that applicable limits of the 
safety analyses are met.
    The proposed changes to reference only the Topical Report number 
and title do not alter the use of the analytical methods used to 
determine core operating limits that have been reviewed and approved 
by the NRC. This method of referencing Topical Reports would allow 
the use of current Topical Reports to support limits in the COLR 
without having to submit an amendment to [the TS of] the operating 
license. Implementation of revisions to Topical Reports would still 
be reviewed in accordance with 10 CFR 50.59 and where required 
receive NRC review and approval.
    Although the relocation of the cycle-specific parameter limits 
to the COLR would allow revision of the affected parameter limits 
without prior NRC approval, there is no significant effect on the 
probability or consequences of an accident previously evaluated. 
Future changes to the COLR parameter limits could result in event 
consequences which are either slightly less or slightly more severe 
than the consequences for the same event using the present parameter 
limits. The differences would not be significant and would be 
bounded by the existing requirement of TS Section 5.6.5c to meet the 
applicable limits of the safety analyses.
    The cycle-specific parameter limits being transferred from the 
TS to the COLR will continue to be controlled under existing 
programs and procedures. The USAR accident analyses will continue to 
be examined with respect to changes in the cycle-dependent 
parameters obtained using NRC reviewed and approved reload design 
methodologies, ensuring that the transient evaluation of new reload 
designs are bounded by previously accepted analyses. This 
examination will continue to be performed pursuant to 10 CFR 50.59 
requirements ensuring that future reload designs will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Additionally, the proposed changes do 
not allow for an increase in plant power levels, do not increase the 
production, nor alter the flow path or method of disposal of 
radioactive waste or byproducts. Therefore, the proposed changes do 
not change the types or increase the amounts of any effluents 
released offsite.
    [The proposed changes to the expression of the 
f1(I) term, which is in the over temperature 
T inequality, clarifies and corrects the term. Moving the 
terms in a Note to the LCO mode applicability is an administrative 
action.]
    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 are programmatic and administrative in 
nature which do not physically alter safety related systems, nor 
affect the way in which safety related systems perform their 
functions.]
    The proposed changes that retain the minimum limit for RCS total 
flow rate in the TS, and that relocate certain cycle-specific 
parameter limits from the TS to the COLR, thus removing the 
requirement for prior NRC approval of revisions to those parameters, 
do not involve a physical change to the plant. No new equipment is 
being introduced, and installed equipment is not being operated in a 
new or different manner. There are no changes being made to the 
parameters within which the plant is operated, other than their 
relocation to the COLR. There are no setpoints affected by the 
proposed changes at which protective or mitigative actions are 
initiated. The proposed changes will not alter the manner in which 
equipment operation is initiated, nor will the function demands on 
credited equipment be changed. No alteration in the procedures which 
ensure the plant remains within analytical limits is being proposed, 
and no change is being made to the procedures relied upon to respond 
to an off-normal event. As such, no new failure modes are being 
introduced.
    The proposed changes to reference only the Topical Report number 
and title do not alter the use of the analytical methods used to 
determine core operating limits that have been reviewed and approved 
by the NRC. This method of referencing Topical Reports would allow 
the use of current Topical Reports to support limits in the COLR 
without having to submit an amendment to [the TS of] the operating 
license. Implementation of revisions to Topical Reports would still 
be reviewed in accordance with 10 CFR 50.59 and where required 
receive NRC review and approval.
    Relocation of cycle-specific parameter limits has no influence 
or impact on, nor does it contribute in any way to the possibility 
of a new or different kind of accident. The relocated cycle-specific 
parameter limits will continue to be calculated using the NRC 
reviewed and approved methodology. The proposed changes do not alter 
assumptions made in the safety analysis and operation within the 
core operating limits will continue.
    [The proposed changes to the expression of the 
f1(I) term, which is in the over temperature 
T inequality, clarifies and corrects the term.]
    Therefore, the proposed changes do not create a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes [are programmatic and 
administrative in nature and] do not physically alter safety related 
systems, nor does it [a]ffect the way in which safety-related 
systems perform their functions. The setpoints at which protective 
actions are initiated are not altered by the proposed changes.
    Therefore, sufficient equipment remains available to actuate 
upon demand for the purpose of mitigating an analyzed event. As the 
proposed changes to relocate cycle-specific parameter limits to the 
COLR will not affect plant design or system operating parameters, 
there is no detrimental impact on any equipment design parameter, 
and the plant will continue to operate within prescribed limits.
    The development of cycle-specific parameter limits for future 
reload designs will continue to conform to NRC reviewed and approved 
methodologies, and will be performed pursuant to 10 CFR 50.59 to 
assure that plant operation [is] within cycle-specific parameter 
limits.
    The proposed changes to reference only the Topical Report number 
and title do not alter the use of the analytical methods used to 
determine core operating limits that have been reviewed and approved 
by the NRC. This method of referencing Topical Reports would allow 
the use of [the] current Topical Reports to support limits in the 
COLR without having to submit an amendment to [the TS of] the 
operating license. Implementation of revisions to Topical Reports 
would still be reviewed in accordance with 10 CFR 50.59 and where 
required receive NRC review and approval.
    [The proposed changes to the expression of the 
f1(I) term, which is in the over temperature 
T inequality, clarifies and corrects the term. Moving the 
terms in a Note to the LCO mode applicability is an administrative 
action.]
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was

[[Page 5344]]

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].

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

    Date of application for amendment: July 9, 2001.
    Brief description of amendment: The amendment revised the 
Administrative Controls Section of the Technical Specifications to 
provide consistency with the changes to Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.59, which were published in the 
Federal Register on October 4, 1999 (64 FR 53582). Specifically, the 
amendment replaced the term ``safety evaluation'' with ``10 CFR 50.59 
evaluation'' and the term ``unreviewed safety question'' with 
``requires NRC [Nuclear Regulatory Commission] approval pursuant to 10 
CFR 50.59.''
    Date of issuance: January 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 239.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44162).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 22, 2002.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: October 31, 2001.
    Brief description of amendment: The amendment deletes Technical 
Specification 5.5.3, ``Post-Accident Sampling,'' eliminating the 
requirement to have and maintain the Post-Accident Sampling System at 
H. B. Robinson. The amendment also deletes Condition 3.G.(4) of the 
Operating License.
    Date of issuance: January 14, 2002.
    Effective date: January 14, 2002.
    Amendment No. 192.
    Facility Operating License No. DPR-23. Amendment revises the 
License and Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64286) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 14, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 30, 2001.
    Brief description of amendment: The amendment deletes requirements 
from the Technical Specifications (and, as applicable, other elements 
of the licensing bases) to maintain a Post-Accident Sampling System.
    Date of issuance: January 14, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 108.
    Facility Operating License No. NPF-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64287).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 14, 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: April 11, 2001, as supplemented 
on September 26 and November 16, 2001.
    Brief description of amendment: The amendment approves a change to 
Technical Specifications 1.12, ``Core Alteration;'' 3.9.1, ``Refueling 
Operations--Boron Concentration;'' 3.9.2, ``Refueling Operations--
Instrumentation;'' and 3.9.11, ``Refueling Operations--Water Level--
Reactor Vessel.'' The amendment also revises the Technical 
Specifications Bases to reflect the changes to the definitions.
    Date of issuance: January 11, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 263.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31705).
    The September 26 and November 16, 2001, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand 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 11, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 23, 2001, as 
supplemented December 20, 2001.
    Brief description of amendment: The amendment revised Technical 
Specification Surveillance Requirement 3.8.4.1 to support replacement 
of the station batteries. The amendment will allow for separate 
required terminal voltage values for the new 31 and 32 station 
batteries.
    Date of issuance: January 17, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 209.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.

[[Page 5345]]

    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59503).
    The December 20, 2001, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 12, 2001.
    Brief description of amendments: The amendments delete TS 6.8.3 
requiring a program for post accident sampling, and thereby eliminate 
the requirements to have and maintain Post Accident Sampling System at 
Donald C. Cook Nuclear Plant, Units 1 and 2.
    Date of issuance: January 16, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 261--Unit 1, 244--Unit 2.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64295)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 16, 2002.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: November 19, 2001.
    Brief description of amendment: The amendment would revise the 
action statement for Technical Specification (TS) 3.3.3.5, ``Remote 
Shutdown Instrumentation,'' to add a statement that the provisions of 
TS 3.0.4 are not applicable.
    Date of issuance: January 16, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 262.
    Facility Operating License No. DPR-58: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64295).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 22, 2001.
    Brief description of amendment: The amendment revised the KNPP TS 
6.14, ``Post Accident Sampling and Monitoring,'' and thereby eliminate 
the requirements to have and maintain the Post Accident Sampling 
System. Although TS 6.14's title contains the word ``monitoring,'' 
elimination of this TS does not eliminate the post-accident monitoring 
instrumentation from KNPP TS. These instruments are contained in KNPP 
TS section 3.5, which are listed in TS Table 3.5-6, ``Accident 
Monitoring Instrumentation Operating Conditions for Indication.''
    Date of issuance: January 16, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 160.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64299).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 30, 2001.
    Brief description of amendment: The amendment eliminates local 
suppression pool temperature limits from the Updated Safety Analysis 
Report as the basis for limiting suppression pool mechanical loads due 
to unstable steam condensation during safety relief valve actuations.
    Date of issuance: January 18, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 126.
    Facility Operating License No. DPR-22. Amendment revised the 
licensing basis.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34286).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 18, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 15, 2001, as supplemented by letters 
dated June 14 and November 21, 2001.
    Brief description of amendment: The amendment: (1) replaced the 
titles of Manager--Fort Calhoun Station and Vice President with generic 
titles, (2) relocated the requirements for the Plant Review Committee 
(PRC) and the Safety Audit and Review Committee (SARC) to the Fort 
Calhoun Station Quality Assurance Program, (3) relocated the 
requirements for procedure controls and records retention to the Fort 
Calhoun Station Quality Assurance Program, (4) enhanced and clarified 
the qualification and training requirements for individuals who perform 
licensed operator functions, (5) incorporated the Westinghouse/CENP 
definition of azimuthal power tilt, and (6) eliminated specific mailing 
address and reporting requirements that are redundant to Title 10 of 
the Code of Federal Regulations.
    Date of issuance: January 11, 2002.
    Effective date: January 11, 2002, and shall be implemented within 
30 days from the date of issuance.
    Amendment No.: 202.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34287). The June 14 and November 21, 2001, supplemental letters 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated January 11, 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: April 2, 2001.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TSs) to relocate TS Sections 3/4.9.4, 
``Refueling Operations,

[[Page 5346]]

Decay Time;'' 3/4.9.5, ``Refueling Operations, Communications;'' 3/
4.9.6, ``Refueling Operations, Refueling Platform;'' and 3/4.9.7, 
``Refueling Operations, Crane Travel--Spent Fuel Storage Pool'' and the 
associated TS Bases pages to the Hope Creek Generating Station Updated 
Final Safety Analysis Report.
    Date of issuance: January 17, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 137.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 16, 2001 (66 FR 
27177).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 25, 2001.
    Brief description of amendment: The amendment deletes Technical 
Specification Section 5.5.3, ``Post Accident Sampling Program'', and 
thereby eliminates the requirements to have and maintain the Post-
Accident Sampling System.
    Date of issuance: January 17, 2002.
    Effective date: January 17, 2002.
    Amendment No.: 81.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64300).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 24, 2001, as 
supplemented by e-mail dated November 16, 2001.
    Brief description of amendments: The amendments decrease the 
calculated peak containment internal pressure for the design basis 
loss-of-coolant accident and main steamline break from 55.1 to 45.9 
psig and 56.6 to 56.5 psig, respectively, in Section 5.5.2.15, 
``Containment Leakage Rate Testing Program,'' of the Technical 
Specifications.
    Date of issuance: January 24, 2002.
    Effective date: January 24, 2002, to be implemented within 60 days 
of issuance.
    Amendment Nos.: Unit 2--182; Unit 3--173.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 3. 2001 (66 FR 
50472).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 24, 2002.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: May 24, 2001.
    Brief description of amendment: This amendment revises Technical 
Specifications Sections 4.2.2.2.e and g, and 4.2.2.4.e and g to adopt a 
modified methodology that relocates the heat flux hot channel factor, 
FQ(z), penalty for increasing FQ(z) versus burnup 
to a table in the Core Operating Limits Report. The amendment also 
increases the surveillance region of FQ(z) to be consistent 
with the current core design and provide assurance that the peak 
FQ(z) is monitored and evaluated near end of core life.
    Date of issuance: January 24, 2002.
    Effective date: January 24, 2002.
    Amendment No.: 153.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38766).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 24, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 9, 2001.
    Brief description of amendments: The amendments consist of changes 
to the Technical Specifications, extending the emergency core cooling 
system accumulator's allowable outage time from 12 hours to 24 hours.
    Date of issuance: January 10, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days from the date of issuance.
    Amendment Nos.: Unit 1--135; Unit 2--124.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44176).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 10, 2002.
    No significant hazards consideration comments received: No.

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: October 31, 2001.
    Brief description of amendments: The amendments delete the program 
requirements of Technical Specification 6.8.4.e, ``Post Accident 
Sampling.''
    Date of issuance: January 14, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 272 and 261.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TSs.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64302).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 14, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 31, 2001.
    Brief description of amendment: The amendment deletes the program 
requirements of Technical Specification 5.7.2.6, ``Post Accident 
Sampling System.''
    Date of issuance: January 14, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 34.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64304).

[[Page 5347]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 14, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 14, 2001.
    Brief description of amendment: The amendment incorporates part of 
TSTF-51, Revision 2 into the Watts Bar Technical Specifications (TS). 
TSTF-51 allows revising the TS to eliminate engineered safety features 
operability requirements that do not involve the movement of irradiated 
fuel during core alterations.
    Date of issuance: January 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
prior to entering Mode 6 for the Cycle 4 refueling outage.
    Amendment No.: 35.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38768).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 14, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification section 3.3.5 ``Loss of Power (LOP) Diesel Generator 
Start Instrumentation,'' to increase the time delay setting of the 
6.9kV shutdown board degraded voltage relays from a nominal 6 seconds 
to 10 seconds.
    Date of issuance: January 23, 2002.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup following the Cycle 4 refueling outage.
    Amendment No.: 36.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38767).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 23, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 2, 2001.
    Brief description of amendments: The amendments delete Technical 
Specification (TS) 5.5.3, ``Post Accident Sampling System,'' and 
thereby eliminate the requirements to have and maintain the Post 
Accident Sampling System (PASS) at Comenche Peak Steam Electric 
Station. In addition, the amendments revise TS 5.5.2, ``Primary Coolant 
Sources Outside Containment,'' to reflect the elimination of PASS.
    Date of issuance: January 15, 2002.
    Effective date: As of the date of issuance and shall be implemented 
by March 15, 2003.
    Amendment Nos.: 91 and 91.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 15, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 29th 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. 02-2567 Filed 2-4-02; 8:45 am]
BILLING CODE 7590-01-P