[Federal Register Volume 67, Number 151 (Tuesday, August 6, 2002)]
[Notices]
[Pages 50947-50965]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-19420]


-----------------------------------------------------------------------

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, July 12, 2002, through July 25, 2002. The 
last biweekly notice was published on July 23, 2002 (67 FR 48213).

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

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

[[Page 50948]]

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 September 5, 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,\1\ which is 
available 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 System's (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.
---------------------------------------------------------------------------

    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest .
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
---------------------------------------------------------------------------

    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. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent 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).

[[Page 50949]]

    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 
System's (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].

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

    Date of amendment request: June 26, 2002.
    Description of amendment request: The licensee proposed to amend 
the Oyster Creek Nuclear Generating Station (OCNGS) Technical 
Specifications (TSs) regarding the safety limit minimum critical power 
ratio (SLMCPR) to reflect the results of cycle-specific calculations 
performed for the next fuel cycle (i.e., Cycle 19), using Nuclear 
Regulatory Commission (NRC)-approved methodology for determining SLMCPR 
values. Specifically, the licensee proposed to revise TS 2.1.A, 
changing the SLMCPR from 1.09 to 1.12 for three-recirculation-loop 
operation, and to 1.11 for four-or five-recirculation-loop operation. 
The proposed amendment would also editorially revise references to 
topical reports which document the approved methodology, and make 
editorial corrections to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The licensee used NRC-approved methods and procedures in Topical 
Report NEDE-24011-P-A-14, ``General Electric Standard Application 
for Reactor Fuel'' (GESTAR II) and U.S. Supplement, NEDE-24011-P-A-
14-US, dated June 2000, to derive the SLMCPR values for OCNGS, Cycle 
19. The analysis methodology incorporates cycle-specific parameters. 
These calculations do not change the operating procedures of OCNGS 
and have no effect on the probability of an accident initiating 
event or transient. The basis of the SLMCPR is to ensure no 
mechanistic fuel damage is calculated to occur if the limit is not 
violated. The new SLMCPR values preserve the existing margin to 
transition boiling and the probability of fuel damage is not 
increased (i.e., in the event of an accident or transient, the 
amount of fuel damaged would not be increased as a result of the new 
SLMCPR values). Furthermore, the proposed new SLMCPR values do not 
lead to, nor do they arise as a result of, plant design or 
procedural changes. The balance of the changes is purely 
administrative. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The new SLMCPR values for OCNGS Cycle 19 core have been 
calculated in accordance with the methods and procedures described 
in NRC-approved topical reports. The proposed new SLMCPR values do 
not lead to, nor do they arise as a result of, plant design or 
procedural changes. The balance of the changes is purely 
administrative. The changes do not involve any new method for 
operating the facility and do not involve any facility 
modifications. As a result, no new initiating events or transients 
could develop from the proposed changes. Therefore, the proposed TS 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The margin of safety as defined in OCNGS's licensing basis will 
remain the same. The new, cycle-specific SLMCPR values are 
calculated using NRC-approved methods and procedures that are in 
accordance with the current fuel design and licensing criteria. The 
SLMCPR values will remain high enough to ensure that greater than 
99.9% of all fuel rods in the core are expected to avoid transition 
boiling if the limits are not violated, thereby preserving the fuel 
cladding integrity. Therefore, the proposed TS changes do not 
involve a significant reduction in a margin of safety.

    Based on the above review, it appears that the three standards of 
10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the requested amendment involves no significant hazards 
consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, 
North Carolina

    Date of amendments request: June 26, 2002.
    Description of amendments request: The proposed amendment would 
revise the Technical Specifications (TS) to revise the reactor coolant 
system pressure-temperature limit curves for operation to 32 effective 
full-power years (EFPY).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Development of the revised BSEP, Unit 1 and 2 
pressure-temperature limits was performed using the approved 
fracture toughness methodologies of 10 CFR 50, Appendix G; the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code, Section XI, Appendix G; and ASME Code Case N-640, 
``Alternative Reference Fracture Toughness for Development of P-T 
Limit Curves for ASME Section XI, Division 1.'' The revised 
pressure-temperature limits were also developed using NRC Regulatory 
Guide 1.190, ``Calculational and Dosimetry Methods for Determining 
Pressure Vessel Neutron Fluence,'' March 2001, for evaluating 
neutron fluence and NRC Regulatory Guide 1.99, Revision 2, 
``Radiation Embrittlement of Reactor Vessel Materials,'' for 
evaluating predicted irradiation effects on vessel beltline 
materials. Use of these methods provides compliance with the intent 
of 10 CFR 50, Appendix G, and provides adequate protection against 
nonductile-type fractures of the reactor pressure vessel. Therefore, 
the probability of occurrence of a previously analyzed event is not 
significantly increased.
    The consequences of a previously evaluated accident are 
dependent on the initial conditions assumed for the analysis, the 
behavior of the fuel during the accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the accident, and the setpoints at which these actions 
are initiated. The proposed revisions do not impact the source term 
or pathways assumed in accidents previously evaluated. No analysis 
assumptions are violated, and there are no adverse effects on the 
factors contributing to offsite and onsite dose. The proposed 
changes to the pressure-temperature limits curves do not affect the 
performance of any equipment used to mitigate the consequences of a 
previously evaluated accident. Also, the proposed changes do not 
affect setpoints that initiate protective or mitigative actions. 
Based on the above, the proposed changes to the pressure-temperature 
limits curves do not significantly increase the consequences of a 
previously evaluated accident.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The changes extend the pressure-temperature limits for use up to 
32 EFPY of

[[Page 50950]]

operation while providing adequate protection against a nonductile-
type fracture of the reactor pressure vessel. Creation of the 
possibility of a new or different kind of accident would require the 
creation of one or more new precursors of that accident. New 
accident precursors may be created by modifications of the plant 
configuration, including changes in allowable modes of operation. 
This proposed license amendment does not involve any facility 
modifications, and plant equipment will not be operated in a 
different manner. Also, no new initiating events or transients 
result from the pressure-temperature limits curves changes. As a 
result, no new failure modes are being introduced. Therefore, the 
proposed changes to the pressure-temperature limits curves will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The margin of safety is established through the design of the 
plant structures, systems, and components; through the parameters 
within which the plant is operated; through the establishment of 
setpoints for actuation of equipment relied upon to respond to an 
event; and through margins contained within the safety analyses. The 
proposed changes to the pressure-temperature limit curves do not 
adversely impact the performance of plant structures, systems, 
components, and setpoints relied upon to respond to mitigate an 
accident. The revised pressure-temperature limits were developed 
using the approved fracture toughness methodologies of 10 CFR 50, 
Appendix G; the American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel Code, Section XI, Appendix G; and ASME 
Code Case N-640, ``Alternative Reference Fracture Toughness for 
Development of P-T Limit Curves for ASME Section XI, Division 1.'' 
The proposed changes are acceptable because the ASME guidance 
maintains the relative margin of safety commensurate with that which 
existed at the time that the ASME Boiler and Pressure Vessel Code, 
Section XI, Appendix G, was approved in 1974. In addition, the 
revised pressure-temperature limits were also developed using NRC 
Regulatory Guide 1.190, ``Calculational and Dosimetry Methods for 
Determining Pressure Vessel Neutron Fluence,'' March 2001, for 
evaluating neutron fluence and NRC Regulatory Guide 1.99, Revision 
2, ``Radiation Embrittlement of Reactor Vessel Materials'' for 
evaluating predicted irradiation effects on vessel beltline 
materials. Use of these methods has provided revised pressure-
temperature limit curves that will ensure that the reactor pressure 
vessel materials continue to behave in a non-brittle manner, thereby 
preserving the original safety design bases[.] No plant safety 
limits, setpoints, or design parameters are adversely affected by 
the proposed changes to the pressure-temperature limit curves. 
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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Kahtan Jabbour, Acting.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, 
North Carolina

    Date of amendments request: July 2, 2002.
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TS) to change the administrative 
controls of TS 5.7, ``High Radiation Area.'' The proposed changes would 
be consistent with the guidance of Regulatory Guide 8.38, ``Control of 
Access to High and Very High Radiation Areas in Nuclear Power Plants,'' 
Section C, Regulatory Position 2.4, Alternative Methods for Access 
Control, with the exception that ``should'' would be changed to 
``shall.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The changes are administrative and affect personnel access 
control requirements for high radiation areas. The changes do not 
affect the operation, physical configuration, or function of plant 
equipment or systems. The changes do not impact the initiators or 
assumptions of analyzed events; nor do they impact the mitigation of 
accidents or transient events. Therefore, these changes do not 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The changes are administrative and affect personnel access 
control requirements for high radiation areas. The changes do not 
alter plant configuration, require installation of new equipment, 
alter assumptions about previously analyzed accidents, or impact the 
operation or function of plant equipment or systems. Therefore, 
these changes will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The changes are administrative and affect personnel access 
control requirements for high radiation areas. The changes do not 
impact any safety assumptions; nor do the changes have the potential 
to reduce any margin of safety as described in the BSEP TS Bases. 
The proposed changes maintain an equivalent level of protection for 
radiation workers and, thereby, provide reasonable assurance that 
individuals will not exceed regulatory dose limits. The proposed 
changes are consistent with: (1) the guidance of Regulatory Guide 
(RG) 8.38, ``Control of Access to High and Very High Radiation Areas 
in Nuclear Power Plants,'' Section C, Regulatory Position 2.4, 
Alternative Methods for Access Control, with the exception that 
``should'' has been changed to ``shall''; (2) the BSEP TSs prior to 
conversion to Improved Standard Technical Specifications; and (3) 
other nuclear plants' existing TSs, including the Crystal River, H. 
B. Robinson, and Shearon Harris nuclear plants.

    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: Kahtan Jabbour, Acting.

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 amendment request: July 8, 2002.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3/4.8.1.1, ``Electrical Power Systems--
A.C. Sources--Operating'' and TS 3/4.8.1.2, ``Electrical Power Systems-
-A.C. Sources--Shutdown'' by revising the minimum level to a volume-
based indication versus a level-based indication.
    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The Harris Nuclear Plant (HNP) Technical Specification (TS) 
Bases for Electrical Power Systems--A. C. Systems states that; ``A

[[Page 50951]]

separate day tank containing a minimum of 1457 gallons of fuel, 
which is equivalent to a minimum indicated level of 40% * * *'' and, 
the asterisked note states; * * * Minimum indicated level with a 
fuel oil specific gravity of 0.83 and the level instrumentation 
calibrated to a reference specific gravity of 0.876.'' These changes 
do not modify the design or operation of Structures, Systems, and 
Components (SSCs) that could initiate an accident. The minimum 
volume of fuel in the day tank is unchanged by this amendment and 
consequently would not impact the probability or consequences of any 
accident scenario.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve new plant components or 
procedures, but only revise existing Technical Specification 
Limiting Condition for Operation Requirements. No significant impact 
on any postulated accident is made due to this change since the 
required fuel oil volume is not changed and the level indication for 
the operations personnel is not changed. These changes do not modify 
the design or operation of Structures, Systems, and Components 
(SSCs) that could initiate an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes do not affect the design or operation of 
safety related components relied upon to automatically mitigate the 
consequences of a design basis event. The day tank level specified 
in TS is not accurate for all fuel oil specific gravities so these 
changes provide better monitoring capability by reducing the 
possibility of confusion. Indicated day tank level is used to 
determine volume by comparing the indicated level to the day tank 
curve using the actual specific gravity of the fuel. The Diesel 
Generator day tank minimum volume is not altered by these changes 
and therefore there * * * is no significant impact on any safety 
system and these changes do not reduce the margin of safety.
    Based on these considerations, the proposed change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Kahtan N. Jabbour, 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: July 11, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to make several administrative 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

First Standard

    Would implementation of this amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. This license amendment request makes editorial corrections 
to several Oconee Technical Specifications. These corrections are 
solely administrative in nature. The deletion of the Reactor 
Building Engineered Safeguards Channels, as proposed in the change 
to the Technical Specification 3.3.6, Engineered Safeguards 
Protective System Manual Initiation, was investigated through Duke's 
corrective action program and also confirmed to be administrative in 
nature. Therefore, all the changes contained in this license 
amendment request are administrative in nature and have no impact on 
any accident probabilities or consequences.

Second Standard

    Would implementation of this amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. There are no new accident causal mechanisms created as a 
result of the implementation of this license amendment request. No 
changes are being made to the plant which will introduce any new 
accident causal mechanisms. This amendment request only makes 
administrative changes and does not impact any plant systems that 
are accident initiators; therefore, no new accident types are being 
created.

Third Standard

    Would implementation of this statement involve a significant 
reduction in a margin of safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. The changes proposed in 
this license amendment request are administrative in nature and do 
not affect the performance of the barriers. Consequently, no safety 
margins will be impacted.

    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: John A. Nakoski.

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

    Date of amendment request: January 10, 2002.
    Description of amendment request: Energy Northwest is requesting 
changes to the technical specifications (TS) to reflect the application 
of a 24-month surveillance test interval (STI) to coincide with its 
intention to implement a 24-month fuel cycle.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The extension of the intervals to 24 months for the subject SRs 
[surveillance requirements] does not impact the ability of any of 
the equipment to function as assumed in the Columbia Generating 
Station accident analysis. None of the equipment within the scope of 
analysis for this TS amendment request performs a function in any of 
the systems required for safe shutdown as described in section 7.4 
of the Columbia Generating Station FSAR [Final Safety Analysis 
Report]. Historical maintenance and surveillance data as well as 
projected instrument drift indicate the proposed amendment will not 
affect performance or reliability of the equipment tested to meet 
the requirements of these SRs. Therefore, the extension of the 
surveillance intervals 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.
    An event related to surveillance testing Frequency or 
instruments drifting beyond Allowable Values is not postulated in 
the Columbia Generating Station accident analysis. None of the 
analyses performed for this amendment request indicate an increase 
in the probability of equipment failure resulting from the 
surveillance interval extension. Because all of the equipment 
related to the proposed SR interval extensions is expected to 
function normally during the longer intervals, extending the subject 
SRs does not introduce any new accident initiators.
    Therefore, the operation of Columbia Generating Station in 
accordance with the

[[Page 50952]]

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 proposed amendment to the Technical Specifications will 
extend the intervals at which testing is performed to meet the 
requirements of the selected SRs. The overall effect of the 
extensions on safety is small due to other more frequent testing 
that is performed on the same equipment, projected instrument drift 
that is bounded by the current setpoint analysis, or the existence 
of redundant mechanical or electrical components. Reviews of 
historical surveillance and maintenance records indicate there is no 
evidence of time-related failures. The proposed amendment does not 
impact the performance of any system, structure, or component relied 
upon for accident mitigation. The proposed surveillance interval 
extensions do not impact any safety analysis assumptions or results.
    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, NW., 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: June 24, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) surveillance requirements (SR) 
3.7.7.1 and SR 3.7.7.2. Specifically, SR 3.7.7.1 would be changed to 
require the verification of the city water tank volume rather than city 
water header pressure and increase the SR frequency from 12 hours to 24 
hours. SR 3.7.7.2 would be revised to require all city water header 
isolation valves are open rather than only the one header supply 
isolation valve.
    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: No.
    The current TS surveillance to verify City Water (CW) header 
pressure did not provide assurance that adequate volume of water was 
available in the City Water Tank (CWT) as an alternate source of 
cooling if Condensate Storage Tank (CST) was not available. The CST 
is not designed to withstand the effect of a tornado-generated 
missile. However, the Auxiliary Feedwater System (AFS) is provided 
sufficient redundancy of water supplies such that an alternate 
source of water from the CWT is available in the event the CST is 
damaged by a tornado-generated missile. The proposed amendment to 
verify CWT volume is [ge]360,000 gallons would ensure that adequate 
volume of CW is available in the CWT to cool the RCS [reactor 
coolant system] from 102% rated thermal power to RHR [residual heat 
removal] entry conditions in 10 hours, if the CST is unavailable or 
depleted for any reason. The surveillance frequency for the CWT 
volume is 24 hours. The proposed amendment to change SR 3.7.7.2 to 
include additional isolation valves that are in the flow path from 
CWT to AFS suction would ensure that all applicable isolation valves 
in the flow path are properly positioned. Thus, the proposed 
amendment involves changes to the Technical Specifications that 
would properly reflect the Surveillance Requirements for CWT. The 
CWT is not an initiator of any accident addressed in the FSAR [Final 
Safety Analysis Report] and the proposed amendment does not have any 
change to the accident analysis addressed in the FSAR.
    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: No.
    The proposed amendment involves changes to the Technical 
Specifications to properly reflect the surveillance requirements of 
City Water Tank. The proposed change provides assurance of 
availability of adequate volume of water in the CWT to cool the RCS 
from 102% rated thermal power to RHR entry conditions in 10 hours, 
if the CST is unavailable or depleted for any reason, and verifies 
the correct position of isolation valves in the flow path between 
the CWT and the AFS pump suction. These changes do not affect any 
accident initiators.
    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: No.
    The proposed amendment involves changes to the Technical 
Specifications to properly reflect the surveillance requirements of 
City Water Tank. The proposed change to verify the CWT volume would 
ensure that an adequate volume of CW is available in the tank to 
cool the RCS from 102% rated thermal power to RHR entry conditions 
in 10 hours, if the CST is unavailable or depleted for any reason. 
The proposed change to verify the valve position for isolation 
valves in the flow path between the CWT and the AFS pump suction 
would ensure that isolation valves in the flow path are properly 
positioned. The proposed amendment does not involve any changes to 
plant equipment, or the way in which the plant is operated.
    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: Richard J. Laufer.

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

    Date of amendment request: June 26, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.6.5.b, ``Core Operating Limits 
Report (COLR),'' to incorporate the reference to Westinghouse topical 
report WCAP-12945-P-A, ``Code Qualification Document for Best Estimate 
Loss-of-Coolant Analysis [LOCA],'' dated March 1998. The proposed 
amendment would also allow the use of the analytical methodology to 
determine the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    No physical changes are being made by this change. The proposed 
changes involve use of the Best Estimate Large Break LOCA [loss-of-
coolant accident] analysis methodology and associated TS [technical 
specification] changes. The plant conditions assumed in the analysis 
are bounded by the design conditions for all equipment in the plant. 
Therefore, there will be no increase in

[[Page 50953]]

the probability of a loss of coolant accident. The consequences of a 
LOCA are not being increased. That is, it is shown that the 
emergency core cooling system is designed so that its calculated 
cooling performance conforms to the criteria contained in 10 CFR 
50.46 paragraph b, that is it meets the five criteria listed in 
Section II of this evaluation. No other accident is potentially 
affected by this change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously analyzed?
    Response: No.
    There are no physical changes being made to the plant. No new 
modes of plant operation are being introduced. The parameters 
assumed in the analysis are within the design limits of existing 
plant equipment. All plant systems will perform equally during the 
response to a potential accident.
    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 proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    It has been shown that the analytic technique used in the 
analysis more realistically describes the expected behavior of the 
Indian Point 3 reactor system during a postulated loss of coolant 
accident. Uncertainties have been accounted for as required by 10 
CFR 50.46. A sufficient number of loss of coolant accidents with 
different break sizes, different locations and other variations in 
properties have been analyzed to provide assurance that the most 
severe postulated loss of coolant accidents were calculated. It has 
been shown by the analysis that there is a high level of probability 
that all criteria contained in 10 CFR 50.46 paragraph b) are met.

    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: Richard J. Laufer.

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

    Date of amendment request: June 26, 2002.
    Description of amendment request: Extend the use of the pressure-
temperature (P-T) limits in Technical Specification (TS) Figure 
3.4.6.1-1 to 32 effective full power years by deleting a note on each 
unit's TS Figure limiting the validity of the Figure.
    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?
    No. The proposed change to the technical specifications to 
extend the use of the existing pressure-temperature (P-T) limits 
does not affect the operation or configuration of any plant 
equipment. Thus, no new accident initiators are created by this 
change. The existing P-T limits are based on the projected reactor 
vessel neutron fluence at 32 effective full power years (EFPY) of 
operation specified in the current licensing basis for LGS [Limerick 
Generating Station], Units 1 and 2. A plant-specific calculation of 
reactor vessel 32 EFPY fast neutron fluence has been completed for 
LGS, Units 1 and 2, using the methodology described in a General 
Electric (GE) Company Licensing Topical Report (LTR), which adheres 
to the guidance in Regulatory Guide 1.190, ``Calculational and 
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.'' 
The three-dimensional spatial distribution of neutron flux was 
modeled by combining the results of two separate two-dimensional 
neutron transport calculations. The latest available cross section 
libraries for the important components of Boiling Water Reactor 
(BWR) neutron flux calculations, i.e., oxygen, hydrogen and 
individual iron isotopes, were included. The resulting reactor 
vessel fast neutron fluence value is lower than the value in the 
current licensing basis for LGS, Units 1 and 2. Therefore, the 
existing 32 EFPY P-T limits bound the fast neutron fluence value 
calculated using the GE methodology. This provides sufficient 
assurance that the LGS, Unit 1 and Unit 2, reactor vessels will be 
operated in a manner that will protect them from brittle fracture 
under all operating conditions. Therefore, the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change to the technical specifications to 
extend the use of the existing P-T limits does not affect the 
operation or configuration of any plant equipment. The current P-T 
limits will remain valid and conservative during the proposed 
extension. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change extends the use of the existing P-T 
limits. The existing P-T limits are based on the projected reactor 
vessel neutron fluence at 32 EFPY of operation specified in the 
current licensing basis for LGS, Units 1 and 2. A plant-specific 
calculation of reactor vessel 32 EFPY fast neutron fluence has been 
completed for LGS, Units1 and 2, using the NRC [Nuclear Regulatory 
Commission] approved methodology in a GE LTR, which adheres to the 
guidance in Regulatory Guide 1.190. The three-dimensional spatial 
distribution of neutron flux was modeled by combining the results of 
two separate two-dimensional neutron transport calculations. The 
latest available cross section libraries for the important 
components of BWR neutron flux calculations, i.e., oxygen, hydrogen 
and individual iron isotopes, were included. The resulting reactor 
vessel fast neutron fluence value is lower than the value in the 
current licensing basis for LGS, Units 1 and 2. Therefore, the 
existing 32 EFPY P-T limits bound the fast neutron fluence value 
calculated using the GE methodology. This provides sufficient margin 
such that the LGS, Unit 1 and Unit 2, reactor vessels will be 
operated in a manner that will protect them from brittle fracture 
under all operating conditions. Therefore, the proposed change does 
not involve a significant reduction in the margin of safety.

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

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
277, Peach Bottom Atomic Power Station, Unit 2, York County, 
Pennsylvania

    Date of application for amendment: June 10, 2002
    Description of amendment request: Exelon Generation Company, LLC, 
the licensee, is proposing a change to the Peach Bottom Atomic Power 
Station (PBAPS), Unit 2, Technical Specifications (TSs) contained in 
Appendix A to the Operating License. This proposed change will revise 
the TS section on safety limits to incorporate revised safety limit 
minimum critical power ratios (SLMCPRs) due to the cycle-specific 
analysis performed by Global Nuclear Fuel for PBAPS, Unit 2, Cycle 15.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 50954]]

issue of no significant hazards consideration, which is presented 
below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The derivation of the cycle specific safety limit minimum 
critical power ratios (SLMCPRs) for incorporation into the (TS[s]), 
and their use to determine cycle specific thermal limits, has been 
performed using the methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which 
incorporates Amendment 25. Amendment 25 was approved by the NRC in a 
March 11, 1999 safety evaluation report.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling. The GE-14 fuel is in compliance with 
Amendment 22 to ``General Electric Standard Application for Reactor 
Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-14-US, June, 2000, which provides the fuel licensing 
acceptance criteria. The probability of fuel damage will not be 
increased as a result of this change. Therefore, the proposed TS 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, calculated to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core if the limit is not violated. The new SLMCPRs are calculated 
using NRC approved methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which 
incorporates Amendment 25. Additionally, the GE-14 fuel is in 
compliance with Amendment 22 to ``General Electric Standard 
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and 
U. S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which provides 
the fuel licensing acceptance criteria. The SLMCPR is not an 
accident initiator, and its revision will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    There is no significant reduction in the margin of safety 
previously approved by the NRC as a result of the proposed change to 
the SLMCPRs, which includes the use of GE-14 fuel. The new SLMCPRs 
are calculated using methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which 
incorporates Amendment 25. The SLMCPRs ensure that greater than 
99.9% of all fuel rods in the core will avoid transition boiling if 
the limit is not violated when all uncertainties are considered, 
thereby preserving the fuel cladding integrity. Therefore, the 
proposed TS change will not involve a significant reduction in the 
margin of safety previously approved by the NRC.
    Based on the above, Exelon Generation Company, LLC, concludes 
that the proposed amendment presents no significant hazards 
consideration under the standards set forth in 10 CFR 50.92(c), and, 
accordingly, a finding of ``no significant hazards consideration'' 
is justified.

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

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: July 18, 2002
    Description of amendment request: The proposed amendments would 
implement an administrative change to relocate the Technical 
Specifications (TS) requirements for the spent fuel crane to the 
respective unit's Updated Final Safety Analysis Report (UFSAR).
    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 operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    No. The proposed changes to the Technical Specifications are 
administrative in nature in that the Technical Specifications for 
operation and surveillance of the spent fuel cask crane and the fuel 
handling crane will be relocated from Appendix A of the facility 
operating license to the UFSAR for each unit. The crane operation 
and surveillance requirements are not altered by this relocation. 
Once relocated, any future changes will be controlled by 10 CFR 
50.59, and the UFSARs will be updated pursuant to 10 CFR 50.71(e). 
Because no operating requirements are changed by the proposed 
amendment, crane operation following the proposed amendment would 
not differ from current crane operation. The proposed Technical 
Specification changes do not involve any change to the configuration 
or method of operation of any plant equipment that is used to 
mitigate the consequences of an accident, nor do the changes alter 
any assumptions or conditions in any of the plant accident analyses. 
Therefore, facility operation in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated in 
the UFSAR.
    2. Would operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    No. The proposed amendment will not affect the design function 
of any system, structure, or component. Relocating the existing 
Technical Specification requirements for the spent fuel cask crane 
and the fuel handling crane to the UFSAR is an administrative change 
and will not modify the physical plant or the modes of plant 
operation defined in the Facility Operating License. The operating 
restrictions imposed on the spent fuel-related cranes by the 
existing Technical Specifications will be retained in the UFSAR 
under this change. The change does not involve the addition or 
modification of equipment, nor does it alter the design or operation 
of plant systems. Therefore, operation of the facility in accordance 
with the proposed amendment would not create the possibility of a 
new or different accident from any accident previously evaluated.
    3. Would operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    No. The proposed changes to the Technical Specifications are 
administrative in nature in that the Technical Specifications for 
operation and surveillance of the spent fuel cask crane and the fuel 
handling crane will be relocated from Appendix A of the facility 
operating license to the UFSAR for each unit. The crane operating 
restrictions that are being relocated to the UFSAR by this change 
are not being relaxed or eliminated. The proposed changes do not 
alter the basis for any technical specification that is related to 
the establishment of or the maintenance of a nuclear safety margin. 
Therefore, operation of the facility in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety as defined in the basis for any Technical Specification or in 
any licensing document.

    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.

[[Page 50955]]

    NRC Section Chief: Kahtan N. Jabbour, Acting.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: July 3, 2002.
    Description of amendment request: The amendment would revise the 
Improved Technical Specifications (ITS) 3.8.1 and associated bases, 
``AC Sources--Operating,'' by extending the allowed outage time for the 
emergency diesel generators (EDGs) from 72 hours to 14 days and to 
modify a note for two EDG ITS Surveillance Requirements (SRs).
    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 not involve a significant increase in the probability 
or consequences of an accident previously analyzed.
    The proposed license amendment extends the Completion Time for 
restoring an inoperable EDG to OPERABLE status and permits 
performance of certain SRs at power under specified conditions. The 
EDGs are designed to supply backup AC power to equipment in 
essential safety systems in the event of a loss of offsite power, 
and as such, the EDGs are not initiators of any design basis 
accident.
    The design functions, operational characteristics, and 
interfaces between the EDGs and other plant systems will not be 
affected by the change. In addition, the initial conditions and 
assumptions for accidents that require the EDGs will remain 
unchanged. Defense in depth will be maintained by the redundant 
OPERABLE EDG, diverse 1E offsite power sources, and the availability 
of multiple emergency feedwater (EFW) and auxiliary feedwater (AFW) 
equipment capable of operating independently of both offsite power 
and the EDGs.
    A Probabilistic Safety Assessment (PSA) has been performed to 
quantitatively assess the risk impact of an increase in Completion 
Times. Although the proposed changes result in slight increases in 
core damage frequency (CDF) and incremental conditional core damage 
probability (ICCDP), and large early release frequency (LERF) and 
incremental conditional large early release probability (ICLERP), 
these increases are well below values that are considered risk 
significant in accordance with current regulatory guidance.
    Based on the above, the proposed changes will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously analyzed.
    The proposed amendment extends the Completion Time for restoring 
an inoperable EDG to OPERABLE status and permits performance of 
certain SRs at power under specified conditions. The proposed 
amendment will not result in changes to the design, physical 
configuration or operation of the plant or the assumptions made in 
the safety analysis for accidents that require the EDGs. In 
addition, the proposed amendment will not result in changes to 
corrective or preventive maintenance activities associated with the 
EDGs, plant operating procedures, or the procedures used to respond 
to abnormal or emergency conditions. Assumptions made in the safety 
analysis related to EDG availability will also remain unchanged. 
Performance of certain SRs at power requires an evaluation to assure 
plant safety is maintained or enhanced, which would include 
evaluation for new or different plant conditions. As such, no new 
failure modes are being introduced. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does not involve a significant reduction in the margin of 
safety.
    The proposed license amendment increases the Completion Times 
for restoring an inoperable EDG to OPERABLE status and permits 
performance of certain SRs at power under specified conditions. The 
proposed changes will improve EDG reliability by providing 
flexibility in scheduling and performing EDG preventive and 
corrective maintenance activities. This flexibility will reduce the 
probability (and associated risk) of a plant shutdown to repair an 
inoperable EDG that cannot be restored within the current ITS 3.8.1 
Completion Times. Performance of the proposed SRs at power requires 
an evaluation to assure plant safety is maintained or enhanced. The 
proposed change will also increase the availability of the EDGs 
during MODE 5 and 6 outages, thus reducing shutdown risk.
    The proposed amendment will not change the plant design, safety 
analysis, or the design, configuration or operation of the EDGs. The 
EDGs are designed to supply backup AC power to equipment in 
essential safety systems in the event of a loss of offsite power. 
Either EDG is capable of performing this function; therefore, as 
long as one train is available, the margin of safety is maintained. 
Defense in depth will be provided by the redundant OPERABLE EDG, the 
availability of diverse offsite circuits capable of supplying power 
to plant emergency loads, and EFW and AFW equipment that can perform 
their design function independently of both offsite power and the 
EDGs.
    To ensure these defense in depth capabilities are maintained 
during required EDG maintenance, maintenance and surveillance 
activities that have the ability to impact the availability of the 
redundant EDG, required support systems and/or backup systems, the 
EFW and AFW systems and the 1E offsite power circuits will be 
controlled in accordance with the normal work controls process. As 
part of this process, weekly qualitative and quantitative risk 
assessments of scheduled on-line maintenance activities, and 
additional risk assessments of emergent work activities, will be 
performed in accordance with the guidance provided in CR-3 
Compliance Procedure CP-253, ``Power Operation Risk Assessment and 
Management.'' If the results of these assessments indicate an 
increase in risk, appropriate actions to control temporary and 
aggregate risk increases and minimize risk increases above the 
overall plant baseline will be implemented in accordance with CP-
253.
    Additional measures to minimize risk will include increased 
administrative controls related to switchyard access, and increased 
inspection of identified risk significant fire areas within the 
plant. A Tier 2 analysis has also been performed to identify the 
dominant risk significant plant configurations during the time that 
an EDG is inoperable due to required corrective or preventive 
maintenance, and appropriate configuration controls/restrictions 
will be established prior to extended EDG maintenance.
    As discussed in question (1) above and in the submittal, the 
slight increases in CDF, ICCDP, LERF and ICLERP resulting from the 
proposed amendment are all below values that are considered risk 
significant in accordance with the guidance provided in Regulatory 
Guide 1.174, ``An Approach for Using Probabilistic Risk Assessment 
in Risk-Informed Decisions on Plant-Specific Changes to the 
Licensing Basis,'' for changes to the plant, and Regulatory Guide 
1.177, ``An Approach for Plant-Specific, Risk-Informed 
Decisionmaking: Technical Specifications,'' for proposed increases 
in ITS Completion Times.
    Based on the above, this proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Acting Section Chief: Kahtan N. Jabbour.

GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear 
Generating Station, Unit 2, (TMI-2) Dauphin County, Pennsylvania

    Date of amendment request: June 13, 2002.
    Description of amendment request: The proposed technical 
specifications change request (TSCR) No. 79, Revision 1, is to revise 
Three Mile Island Nuclear Generating Station, Unit 2 (TMI-2) Technical 
Specification (TS) Administrative Controls section that will provide 
consistency with Three Mile Island Nuclear Generating Station, Unit 1, 
(TMI-1) TS changes submitted

[[Page 50956]]

by AmerGen Energy Company, LLC (AmerGen) and Exelon Generation Company, 
LLC (EGC), which are currently under review by the U.S. Nuclear 
Regulatory Commission (NRC). GPU Nuclear utilizes EGC/AmerGen 
administrative controls under contract to TMI-2. The proposed request 
would delete TS Sections 6.4, ``Training,'' and 6.5.4, ``Independent 
Onsite Safety Review Group'' (IOSRG) from the administrative 
requirements in Section 6 of the TMI-2 Post Defueled Monitored Storage 
(PDMS) TS. Additionally, the IOSRG has been removed from the list of 
recipients of audit reports in Section 6.5.3.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    TMI-2 is a defueled facility holding a Possession Only License 
is being maintained in Post Defueling Monitored Storage (PDMS). The 
introduction of the PDMS Quality Assurance Plan states in part in 
the second paragraph, ``Since the plant will be in a non-operating 
and defueled status, there will no longer be any structures, 
systems, or components that perform a safety function.''
    Deletion of the technical specifications requirements for 
training and the IOSRG will have no adverse effect on any plant 
system; will not alter the source term, containment isolation, or 
allowable radiological consequences. These administrative changes 
will have no effect on any plant systems, structures or components 
and do not affect the physical plant, operating procedures, 
maintenance procedures, or emergency procedures at TMI-2.
    The elimination of the IOSRG oversight function removes a 
function that is redundant to other oversight programs, not required 
by NRC regulation, and is not needed for the safe monitoring of TMI-
2. Programmatic assessments of the TMI-2 programs will continue to 
be assessed by Nuclear Oversight personnel in accordance with the 
PDMS Quality Assurance Plan. Training will continue to be conducted 
in accordance with regulatory requirements.
    The training programs for appropriate unit staff personnel other 
than licensed operators is now addressed by 10 CFR 50.120. With the 
10 CFR 50.120 rule, the NRC is emphasizing the need to ensure that 
industry personnel training programs are based upon job performance 
requirements. This will be accomplished using the systems approach 
to training implemented by INPO [Institute of Nuclear Power 
Operations] accredited training programs for selected nuclear 
personnel. Included within the rule is the requirement that the 
training program must reflect industry experience. Deletion of the 
training requirements in the technical specifications will conform 
the license to the current requirements of 10 CFR 50.120.
    Therefore, these changes 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?
    These changes are administrative in nature and do not affect any 
system functional requirements, plant maintenance, or operability 
requirements. The proposed changes involve the elimination of a 
redundant oversight function and the replacement of training 
requirements by the more vigorous requirements of 10 CFR 50.120, 
which are applicable to operating plants.
    The proposed changes have no direct effect on any plant systems 
or components. The programs for the monitoring, surveillance, or 
maintenance of TMI-2 are unaffected. Oversight of TMI-2 will 
continue to be provided by Nuclear Oversight personnel and the TMI-2 
Safety Oversight Committee in accordance with the requirements of 
the PDMS Quality Assurance Plan.
    Therefore, the proposed changed will 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 training and IOSRG requirements contained in TMI-2 Technical 
Specifications Section 6.0 ``Administrative Controls'' are 
administrative in nature. The proposed changes have no direct effect 
on any plant systems. There are currently no safety limits that 
apply to TMI-2 during PDMS. 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 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 & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: June 28, 2002.
    Description of amendment request: The licensee proposed changes to 
surveillance requirements in Table 4.6.2b, ``Instrumentation that 
Initiates Primary Coolant System or Containment Isolation,'' of the 
Nine Mile Point Nuclear Station, Unit No. 1 (NMP1) Technical 
Specifications (TS) regarding the isolation capability of the shutdown 
cooling system (SDCS). Specifically, the changes will remove the 
restriction to perform channel functional testing and channel 
calibration associated with SDCS high area temperature only during 
refueling outages. The changes will allow these surveillance activities 
to be performed during other operating conditions on a once-per-
operating-cycle basis, thereby maintaining SDCS availability to support 
reactor shutdown operations during refueling.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The operation of Nine Mile Point Unit 1 in accordance with 
the proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The only safety-related functions of the SDCS are (i) to 
maintain the integrity of the reactor coolant pressure boundary, and 
(ii) to provide primary containment isolation of the shutdown 
cooling lines. The proposed amendment removes an unnecessary 
restriction to perform channel functional testing and calibration 
associated with SDCS isolation capability only during refueling 
outages. It provides the flexibility to perform these surveillances 
during other operating conditions on a ``once per operating cycle'' 
basis. The change does not modify the surveillance frequency, 
surveillance acceptance criteria, high area temperature setpoint 
limit for initiating SDCS isolation, plant equipment configurations 
during SDCS surveillances, or the existing requirements for 
maintaining SDCS isolation and reactor coolant pressure boundary 
integrity.
    Based on the above, the operation of NMP1 in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or the consequences of an accident previously evaluated.
    2. The operation of Nine Mile Point Unit 1 in accordance with 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not involve any physical modifications 
to the plant and does not alter equipment configuration, setpoints, 
safety parameters, surveillance interval durations, or surveillance 
acceptance criteria. It does not affect the operation of any safety-
related structure, system, or component in a manner that could 
introduce a new accident precursor or a new failure mechanism. The 
SDCS isolation valves will continue to perform their isolation 
function by remaining closed with power removed during power 
operation of the reactor.
    Based on the above, the operation of NMP1 in accordance with the 
proposed amendment cannot create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The operation of Nine Mile Point Unit 1 in accordance with 
the proposed

[[Page 50957]]

amendment will not involve a significant reduction in a margin of 
safety.
    The proposed change does not affect any of the plant's fission 
product barriers or safety/operational limits. The high area 
temperature setpoint for SDCS isolation will remain within the 
existing TS limit.
    The SDCS isolation valves will continue to remain closed with 
power removed during power operation of the reactor. The proposed 
``[o]nce per operating cycle'' surveillances will be adequate to 
ensure acceptable SDCS equipment operability and reliability.
    Based on the above, the operation of NMP1 in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.

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

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: July 12, 2002.
    Description of amendment request: The licensee proposed to change 
the Technical Specifications (TSs), Sections 3.1.1 and 4.1.1, ``Control 
Rod System,'' by reducing the power level below which the rod worth 
minimizer (RWM) or a second independent verification of rod positions 
must be used from 20% rated thermal power (RTP) to 10% RTP. The 
licensee stated that analysis has shown that no significant control rod 
drop accident (CRDA) can occur above 10% RTP. The low power setpoint 
change will reduce the time necessary for both reactor startup and 
shutdown.
    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 reproduced below:

    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The TS revision lowers the power level at which the analyzed rod 
position sequence must be followed by use of the RWM or a second 
independent verification of rod positions. The RWM enforces the 
analyzed rod position sequence to ensure that the initial conditions 
of the CRDA analysis are not violated. Compliance with the analyzed 
rod position sequence and operability of the RWM is required in the 
startup and run modes when thermal power is less than 10% RTP. When 
thermal power is 10% RTP or greater, there is no possible control 
rod configuration that results in a control rod worth that could 
exceed the 280 cal/gram fuel design limit during a CRDA. None of the 
accidents previously evaluated assume the RWM is an initiator of the 
accident and therefore, the probability of an accident is not 
significantly increased by the change. Because the fuel design limit 
is not exceeded, the change to the low power setpoint will not 
significantly increase the consequences of an accident previously 
evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The TS revision lowers the power level below which the analyzed 
rod position sequence must be followed. The change does not 
introduce a new mode of plant operation and does not involve a 
physical modification to the plant. Therefore, a new or different 
type of accident from any accident previously evaluated is not 
created.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The RWM enforces the analyzed rod position sequence to ensure 
that the initial conditions of the CRDA analysis are not violated. 
Compliance with the analyzed rod position sequence and operability 
of the RWM are required in the startup and run modes when thermal 
power is less than 10% RTP. When thermal power is 10% RTP and 
greater, there is no possible control rod configuration that results 
in a control rod worth that could exceed the 280 cal/gram fuel 
design limit during a CRDA. Because the fuel design limit is not 
exceeded at 10% RTP and greater, the change to the RWM low power 
setpoint does not significantly reduce 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: July 12, 2002.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) Section 3.1.a.3, ``Pressurizer Safety Valves.'' Also, the proposed 
amendment would reformat TS 3.1.a.3 to more closely resemble the format 
of Improved Standard Technical Specification (ISTS) to improve clarity. 
The proposed amendment would allow both pressurizer safety valves to be 
inoperable or removed while the reactor vessel head is on. This would 
only be applicable when the temperature and pressure are low enough 
such that the Low Temperature Overpressure Protection (LTOP) System can 
safely protect the Reactor Coolant System (RCS). The TSs currently 
requires the LTOP System to protect the RCS when the RCS temperature is 
less than LTOP enabling temperature.
    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 format changes are administrative in nature and therefore 
have no effect on the probability or consequences of an accident. 
The situation where the plant has two inoperable or removed 
pressurizer safeties while the LTOP System is enabled is not 
considered an accident initiator. Therefore, any change to the 
system would not affect the probability of an accident previously 
evaluated. The risk of core damage/release of radioactivity would 
not increase with all of the other plant safety features still in 
place.
    The proposed changes adds clarity to the TSs by describing a 
specific situation when the RCS is at low temperature & pressure 
while overpressure protection is provided by the LTOP System. Since 
this TS change is not an accident initiator and existing TS will 
ensure the LTOP System will continue to protect the RCS pressure 
boundary, this proposed amendment does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The situation where the plant has two inoperable pressurizer 
safeties while the LTOP System is enabled is not considered an 
accident initiator. A failure of this system will not result in an 
accident. The format changes are administrative in nature and 
therefore have no effect on the probability or consequences of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a change to the physical 
plant or operations. As the RCS temperature is lowered to less than 
200  deg.F, the LTOP System provides the RCS overpressure protection 
required. Since the LTOP System is currently approved for use by TS 
3.1.b.4, it would not create the

[[Page 50958]]

possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, any change to the system would not affect the 
probability of an accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The format changes are administrative in nature and therefore 
are not involved in a significant reduction in the margin of safety. 
Margin of safety relates to overpressure protection when the RCS is 
less than 200  deg.F. This margin is controlled by the LTOP System 
completely and does not rely on the pressurizer safeties. This 
proposed amendment allows KNPP to have both pressurizer safeties to 
be inoperable as long as the RCS is below the LTOP System enabling 
temperature. Therefore, NMC concludes that there is not a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

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: June 24, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.3, ``Post Accident Sampling 
System (PASS),'' to eliminate the requirements to have and maintain the 
PASS at Plant Hatch. The changes are based on NRC-approved Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-413, ``Elimination of Requirements for a Post Accident 
Sampling System (PASS).''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on December 27, 2001 (66 FR 66949), on possible 
amendments concerning TSTF-413, 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 application in the Federal Register on 
March 20, 2002 (67 FR 13027). The licensee affirmed the applicability 
of the following NSHC determination in its application dated June 24, 
2002.
    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 PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the [Three Mile Island, Unit 2] TMI-2 accident. The specific 
intent of the PASS was to provide a system that has the capability 
to obtain and analyze samples of plant fluids containing potentially 
high levels of radioactivity, without exceeding plant personnel 
radiation exposure limits. Analytical results of these samples would 
be used largely for verification purposes in aiding the plant staff 
in assessing the extent of core damage and subsequent offsite 
radiological dose projections. The system was not intended to and 
does not serve a function for preventing accidents and its 
elimination would not affect the probability of accidents previously 
evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
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 elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the 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 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: John A. Nakoski.

[[Page 50959]]

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: July 11, 2002.
    Description of amendment request: The proposed amendments would 
delete Technical Specification 3.3.1.1.I.2, which requires returning 
the Oscillating Power Range Monitor to operable status within 120 days 
of discovering its operability.
    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 Oscillating Power Range Monitor (OPRM) is not designed for 
the prevention of an instability event or any other previously 
evaluated event. Accordingly, it cannot increase the probability of 
an instability event or any other previously evaluated event.
    The consequences of the instability event are not significantly 
increased, because the alternate method of detection and suppression 
of thermal-hydraulic instability oscillations is well established at 
Plant Hatch. Furthermore, operators are adequately trained on 
instabilities.
    This proposed change to delete the 120-day Completion Time 
restriction on an inoperable OPRM does not affect any other system 
designed for the mitigation of previously analyzed events.
    For the above reasons, the probability and consequences of a 
previously analyzed event are not increased.
    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 only deletes a Technical Specification 
requirement. It does not physically alter the design, operation, 
testing, or maintenance of any plant system or piece of equipment. 
The proposed change introduces no new modes of operation. 
Consequently, the change does not create the possibility of a new or 
different kind [of] event.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed change deletes the requirement to restore the OPRM 
system to operable status within 120 days of discovering its 
inoperability. A manual alternate method to detect and suppress 
thermal-hydraulic instability oscillations has been included in 
Plant Hatch procedures for many years. Also, operators are trained 
on instability events.
    Accordingly, the manual alternate method is adequate and thus, 
the margin of safety for the instability event is not significantly 
reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that 
the amendment request involves no significant hazards consideration.

    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: John A. Nakoski.

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

    Date of amendment request: August 2, 2001.
    Description of amendment request: The proposed amendment revises 
Technical Specifications to extend, on a one-time basis, the current 
interval for Type A testing from 10 years 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:

    The proposed Technical Specification revision extends the 
current interval for Type A testing. The current test interval of 
ten years would be extended on a one-time basis to 15 years from the 
preceding Type A test. Pursuant to 10 CFR 50.91, this analysis 
provides a determination that the proposed change to the Technical 
Specifications for a one-time extension of the interval for 
Integrated Leakage Rate Testing does not involve any significant 
hazards consideration as defined in 10 CFR 50.92.

Criterion 1: The proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.

    The proposed extension to the Type A testing interval will not 
increase the probability of an accident previously evaluated. The 
containment Type A testing interval extension is not a modification 
and the testing interval extension is not of a type that could lead 
to equipment failure or accident initiation.
    The proposed extension to the Type A testing interval does not 
involve a significant increase in the consequences of an accident. 
Research documented in NUREG-1493 has determined that Type B and C 
tests can identify the vast majority (more than 95%) of all 
potential leakage paths.
    NUREG-1493 concluded that reducing the Type A test frequency to 
one per twenty years leads to an imperceptible increase in risk. 
Testing and inspection provide a high degree of assurance that the 
containment will not degrade in a manner detectable only by Type A 
testing. Previous Type A tests show leakage does not exceed 
acceptance criteria, indicating a very leak-tight containment. 
Inspections required by the Maintenance Rule and ASME code are 
performed in order to identify indications of containment 
degradation that could affect leak tightness.
    Experience at the South Texas Project demonstrates that 
excessive containment leakage paths are detected by Type B and C 
Local Leakage Rate Tests. Type B and C testing will identify any 
containment opening, such as a valve, that would otherwise be 
detected by the Type A tests. These factors show that a Type A test 
interval extension will not involve a significant increase in the 
consequences of an accident.

Criterion 2: The proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.

    The proposed extension of the Type A testing interval will not 
create the possibility of a new or different type of accident from 
any previously evaluated. There are no physical changes being made 
to the plant and there are no changes in operation of the plant that 
could introduce a new failure mode creating an accident or affecting 
the mitigation of an accident.

Criterion 3: The proposed change does not involve a significant 
reduction in the margin of safety.

    The proposed extension of the Type A testing interval will not 
significantly reduce the margin of safety. The NUREG-1493 generic 
study of the effects of extending containment leakage testing found 
that a 20-year interval in Type A leakage testing results 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 effect on this risk because 
95% of the potential leakage paths are detected by Type B and C 
testing.
    Deferral of Type A testing for the South Texas Project does not 
increase the level of public risk due to loss of capability to 
detect and measure containment leakage or loss of containment 
structural capability. Other containment testing methods and 
inspections will assure all limiting conditions of operation will 
continue to be met. The margin of safety inherent in existing 
accident analyses is maintained.
    Based on the evaluation provided above, the South Texas Project 
concludes that the proposed change does not involve a significant 
hazards consideration and will not have a significant effect on safe 
operation of the plant. Therefore, there is reasonable

[[Page 50960]]

assurance that operation of the South Texas Project in accordance 
with the proposed revised Technical Specifications will not endanger 
the public health and safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 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: Alvin H. Gutterman, Esqr., Morgan, Lewis, & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of amendment request: July 10, 2002.
    Description of amendment request: The proposed one-time technical 
specification (TS) change revises the Sequoyah Unit 2 Limiting 
Condition for Operation for Section TS 3.7.4, ``Essential Raw Cooling 
Water System,'' to include provisions for maintaining operability of 
this system during performance of heavy load lifts associated with the 
Unit 1 steam generator replacement (SGR) project. The provisions should 
ensure safe operation of Unit 2 during heavy load lift activities. In 
addition, compensatory measures proposed should ensure safe shutdown 
capability of Unit 2 in the unlikely event a heavy load drop occurs 
over Essential Raw Cooling Water system piping.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the Tennessee Valley 
Authority (TVA) has provided its analysis of the issue of no 
significant hazards consideration, which is presented below:

    TVA has concluded that operation of Sequoyah (SQN) Unit 2, in 
accordance with the proposed change to Technical Specification (TS) 
3/4.7.4, does not involve a significant hazards consideration. TVA's 
conclusion is based on its evaluation, in accordance with 10 CFR 
50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).
    TVA's proposed license amendment is a one-time change to the SQN 
Unit 2 TSs. The proposed change revises SQN Limiting Condition for 
Operation 3.7.4, ``Essential Raw Cooling Water System,'' to include 
provisions for maintaining operability of this system during 
performance of heavy load lifts associated with the Unit 1 steam 
generator replacement (SGR) project. The provisions ensure safe 
operation of Unit 2 during heavy load lift activities. In addition, 
compensatory measures ensure safe shutdown capability of Unit 2 in 
the unlikely event a heavy load drop occurs over ERCW system piping.
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    No changes in event classification as discussed in SQN Updated 
Final Safety Analysis Chapter 15 will occur due to the proposed TS 
amendment. The one-time TS provision ensures that the SQN essential 
raw cooling water (ERCW) system remains operable for continued safe 
operation of Unit 2 during heavy load lifts performed on Unit 1 
during SGR replacement activities.
    Accordingly, the proposed modification to SQN Unit 2 TSs and the 
implementation of compensatory measures for a postulated load drop 
will not significantly increase 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 possibility of a new or different accident scenario 
occurring as a result of activities conducted during the SQN Unit 1 
SGR project are not created. Three postulated scenarios related to 
heavy load handling during the SGR project were examined for their 
potential to represent a new or different kind of accident from 
those previously evaluated: (1) a breach of the old steam generator 
(OSG), resulting in the release of contained radioactive material, 
(2) flooding in the Auxiliary Building caused by the failure of 
piping in the ERCW tunnel, and (3) loss of ERCW to support safe 
shutdown of the operating unit.
    Failure of an OSG that results in a breach of the primary side 
of the steam generator (SG) could potentially result in a release of 
a contained source outside containment. The consequences of this 
event, both offsite and in the control room, were examined and found 
to be within the consequences of the failure of other contained 
sources outside containment at the SQN site (i.e., within the SQN 
design basis).
    With regard to flooding of the Auxiliary Building from a heavy 
load drop, the protective measure taken prior to the lifting of 
heavy loads include installation of a wall in the ERCW tunnel near 
the Auxiliary Building interface. The wall provides protection 
against a postulated flood of the ERCW tunnel and protects against 
flooding of the Auxiliary Building beyond those events previously 
evaluated.
    With regard to the potential for a heavy load drop causing the 
loss of ERCW cooling water to the operating unit (i.e., Unit 2), TVA 
is implementing provisions to preclude a load drop. A heavy load 
drop is considered an unlikely accident for the following reasons:
     The lifting equipment was specifically designed and 
chosen for the subject heavy lifts,
     Crane operators will be specially trained in the 
operation of the lift equipment and in the SQN site conditions,
     Qualifying analyses and administrative controls will be 
used to protect the lifts from the effects of external events,
     The areas over which a load drop could cause loss of 
ERCW are a small part of the total travel path of the loads.
    In addition, protection against the potential for a loss of ERCW 
is established prior to any heavy load lifts. Compensatory measures 
ensure the ERCW system is isolated should a pipe break occur, and 
that ERCW flow is redirected to equipment essential for safe 
shutdown capability of Unit 2.
    Accordingly, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to the Unit 2 TSs support safe operation and 
safe shutdown capability of Unit 2 during replacement of the Unit 1 
SGs. These measures do not result in changes in the design basis for 
plant structures, systems, and components (SSCs). Consequently, the 
proposed change will not affect any margins of safety for plant 
SSCs.
    Accordingly, a significant reduction in the margin of safety is 
not created by the proposed change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
    NRC Section Chief: Kahtan N. Jabbour, Acting.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendments: March 29, 2002 (TS 02-02).
    Brief description of amendments: The proposed amendment would 
change the Sequoyah (SQN) Unit 1 Technical Specifications (TSs) by 
revising Specification 3/4.4.5 to eliminate surveillance requirements 
associated with two alternate repair criteria. The associated License 
Condition 2.C.9.d is also deleted. In addition, the proposed change 
revises SR 3/4.4.5.3.a to allow a one-time, 40-month steam generator 
(SG) inspection interval after the first (post-Unit 1 SG replacement) 
inservice inspection resulting in a C-1 category. The proposed change 
is in lieu of the current TS criteria that requires two consecutive 
category C-1 inspections for application of the 40-month SG inspection 
interval.
    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:


[[Page 50961]]


    TVA has concluded that operation of Sequoyah Nuclear Plant (SQN) 
Unit 1, in accordance with the proposed change to the technical 
specifications and License Condition, does not involve a significant 
hazards consideration. TVA's conclusion is based on its evaluation, 
in accordance with 10 CFR 50.91(a)(1), of the three standards set 
forth in 10 CFR 50.92(c).
    TVA is proposing to modify SQN Unit 1 TS 3/4.4.5, ``Steam 
Generators'' to delete surveillance requirements (SRs) that describe 
steam generator (SG) tube plugging limits for two alternate repair 
criteria (ARC). The first ARC is for axial outside diameter stress 
corrosion cracking (ODSCC) at non-dented tube support plates and the 
second ARC is for axial primary water stress corrosion cracking 
(PWSCC) at dented tube support plates. TVA's proposed amendment 
removes both ARCs through the deletion of the following SRs: SR 
4.4.5.2.b.4, 4.4.5.2.d, 4.4.5.2.e, a portion of 4.4.5.4.a.6, 
4.4.5.4.a.10, 4.4.5.4.a.11, 4.4.5.5.d, and 4.4.5.5.e. TVA's proposed 
removal of these SRs for ARC reestablishes standard tube plugging 
criteria within the TS for SQN Unit 1. Returning to the standard TS 
40 percent through-wall tube plugging limit is inherently more 
conservative.
    Included with the above change is deletion of License Condition 
2.C.9.d that references prior TVA commitment letters for SG 
inspection. The TVA letters and their commitments will no longer 
apply following replacement of the Unit 1 SGs.
    In addition, TVA is proposing a revision to TS 3/4.4.5.3.a to 
allow application of the 40-month inspection interval after one SG 
inspection resulting in a C-1 category. The proposed change replaces 
the current TS requirement that invokes the extended 40-month 
inspection interval after two consecutive inspections resulting in a 
category of C-1. TVA's proposed change provides a relaxation of the 
SG inspection requirements and schedule. The relaxation in the 
inspection schedule is intended to coincide with replacement of SQN 
Unit 1 SGs during the Cycle 12 refueling outage (Spring 2003). The 
replacement of the SQN Unit 1 SGs incorporate significant design 
improvements that include thermally treated Alloy 690 SG tubing. The 
improvements in SG design and tube material properties increase the 
resistance to SG tube degradation mechanisms and allow optimization 
of SG inspection schedules. The proposed optimization of SG 
inspections reduce the cumulative number of SG inspections over the 
life of the plant and result in significant dose, schedule, and cost 
savings to TVA.
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    TVA's proposed TS amendment does not compromise limits 
associated with SG tube integrity. TVA's proposed change removes 
existing SG tube plugging criteria (i.e., ARC) from the TS and 
reestablishes the standard TS criteria (40 percent through-wall 
criteria). This change is inherently more conservative. The proposed 
allowance for an extended inspection interval is a conservative 
inspection strategy that is based on improved SG design features and 
SG tube materials that have been shown to resist degradation and 
preserve SG tube integrity.
    The proposed revision does not alter plant equipment, test 
methods or operating practices. The proposed change continues to 
provide controls for safe operation of SQN SGs within the required 
limits. The proposed change does not contribute to events or 
assumptions associated with postulated design basis accidents (i.e., 
SG tube rupture). The proposed change does not affect operator 
indicators or actions required to diagnose or mitigate a SG tube 
rupture accident. The proposed revisions continue to maintain the 
required safety functions. Accordingly, the probability of an 
accident or the consequences of an accident previously evaluated is 
not increased.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    TVA's proposed amendment removes existing repair criteria and 
incorporates the more conservative TS limit for SG tube plugging 
(i.e., plug tubes with degradation depths equal to or greater than 
40 percent through-wall). This change will not give rise to new 
failure modes. The failure of a SG tube to maintain leakage 
integrity during operation is an analyzed event in the SQN Updated 
Final Safety Analysis Report. TVA's proposed change to the SG 
inspection interval will not introduce a new or different kind of 
accident scenario. Accordingly, 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.
    TVA's proposed TS amendment is conservative with respect to the 
margin of safety. The margin of safety is preserved through ensuring 
structural integrity and leakage integrity of the SG tubes.
    TVA's proposed change that to remove ARC from the TS does not 
compromise structural integrity or leakage integrity of SG tubes. 
The proposed change invokes the standard TS tube plugging criteria 
limit (40 percent through-wall criteria) which is inherently 
conservative.
    TVA's proposed change to include a one-time extension to the SQN 
Unit 1 SG inspection interval retains conservative inspection 
strategy that maintains the structural and leakage integrity of the 
SGs. TVA intends to replace SQN Unit 1 SGs during the Cycle 12 
refueling outage and perform a 100 percent full length inspection of 
SG tubes during the Cycle 13 refueling outage to verify that damage 
mechanisms do not exist. Twelve years of SG operation history 
indicate that corrosion damage mechanisms do not appear in 
replacement SGs that contain thermally treated Alloy 690 tubing. The 
replacement SG design also contains design improvements that provide 
reasonable assurance that tube degradation is not likely to occur 
over the proposed 40-month operating period (Cycle 13 refueling 
outage to Cycle 15 refueling outage). The corrosion resistant 
properties of the thermally treated Alloy 690 tubing and the 
improved design will limit the initiation of damage mechanisms and 
limit growth rate such that tube structural and leakage integrity 
will be maintained over two operating cycles.
    TVA's proposed change to extend the SG inspection interval does 
not result in a change to system design features. The proposed 
change does not affect the plant conditions, setpoints, or safety 
limits that could result in precursors to accidents or degrade 
accident mitigation systems. Accordingly, plant system safety 
functions are not altered by the proposed change.
    The effect of this change is to extend allowable SG inspection 
intervals while retaining conservative margins to maintain the 
structural and leakage integrity of the SGs. Consequently, the 
proposed TS revisions does 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: Kahtan N. Jabbour, Acting.

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: July 10, 2002 (TS 01-09).
    Brief description of amendments: The proposed amendment would 
change the Sequoyah (SQN) Unit 1 and 2 Technical Specifications (TSs) 
by removing the requirement to not make positive reactivity changes 
during certain conditions and replace it with requirements to maintain 
shutdown margin or boron concentration. The changes will permit limited 
positive reactivity changes that are necessitated by plant operations. 
These changes will limit the amount of reactivity changes to those that 
will continue to assure appropriate reactivity limits are met. The 
proposed changes are consistent with TS Task Force 286 and Revision 2 
to NUREG-1431.
    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

[[Page 50962]]

probability or consequences of an accident previously evaluated.
    The proposed change does not involve an increase in the 
probability or consequences of an accident previously evaluated. The 
proposed activities to be allowed during certain operating 
conditions are permitted at other times during routine operating 
conditions. The changes do not affect the limits on reactivity that 
are specified in other specifications. The proposed changes continue 
to ensure restrictions on additions and flowpaths of unborated water 
that are in the existing specifications. The proposed change does 
not affect the limits on reactivity that are credited in the safety 
analysis. Therefore, no increase in the probability or consequences 
of any accident previously evaluated will occur.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes permit the conduct of normal operating 
evolutions during limited periods when additional controls over 
reactivity margin are imposed by the TSs. The proposed change does 
not introduce any new equipment into the plant or significantly 
alter the manner in which existing equipment will be operated. The 
changes to operating allowances are minor and are only applicable 
during certain conditions. The operating allowances are consistent 
with those acceptable at other times. Since the proposed changes 
only allow activities that are presently approved and routinely 
conducted, no possibility exists for a new or different kind of 
accident from those previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety because the ability to make the reactor subcritical 
and maintain it subcritical during all operating conditions and 
modes of operation will be maintained. The margin of safety is 
defined by the shutdown margin limits and the refueling boron 
concentration limit. The proposed changes do not affect these 
operating restrictions and the margin of safety which assures the 
ability to make and maintain the reactor subcritical is not 
affected.

    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: Kahtan N. Jabbour, Acting.

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

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

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of amendment request: July 18, 2002.
    Brief description of amendment request: These amendments would 
revise the Facility Operating Licenses (FOLs) to change the 
implementation date for the Improved Technical Specifications (ITS), 
including the relocation of certain existing TS requirements to 
licensee-controlled documents, from no later than September 2, 2002, to 
no later than December 20, 2002.
    Date of publication of individual notice in Federal Register: July 
25, 2002 (67 FR 48679).
    Expiration date of individual notice: August 26, 2002.

Notice of Issuance of Amendments to Facility Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/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. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: December 28, 2000, as 
supplemented May 31, 2002.
    Brief description of amendment: The amendment decreases the allowed 
outage time for an inoperable channel or channels of the anticipated 
transient without scram recirculation pump trip instrumentation.
    Date of issuance: July 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 153.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9378). The supplemental letter did not significantly change the 
requested amendment or affect the proposed no significant hazards 
consideration determination.

[[Page 50963]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 31, 2001, as supplemented by 
letters dated March 6, and April 23, 2002.
    Brief description of amendment: The amendment revises License 
Condition 2.C.(3) of Operating License DPR-6 to reference revisions of 
the Big Rock Point Defueled Security Plan, Defueled Suitability 
Training and Qualification Plan, Defueled Safeguards Contingency Plan, 
and Independent Spent Fuel Storage Installation Security Plan.
    Date of issuance: July 18, 2002.
    Effective date: As of the date of issuance and shall be implemented 
prior to placing the spent fuel in the Big Rock Point Plant independent 
spent fuel storage installation.
    Amendment No.: 123.
    Facility Operating License No. DPR-6: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44166). The March 6 and April 23, 2002, supplemental letters provided 
additional clarifying information that did not expand the scope of the 
application as originally noticed and did not change the NRC staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated July 18, 2002.
    No significant hazards considerations comments received: No.

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

    Date of application for amendment: February 5, 2002 as supplemented 
on March 6, 2002.
    Brief description of amendment: The amendment changes the term in 
the technical specifications ``once each REFUELING INTERVAL'' to ``once 
per 24 months'' in several surveillance requirements.
    Date of issuance: July 24, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 206.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36930). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 24, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 20, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 5.5.14 to eliminate the use of the term 
``unreviewed safety question,'' and replace the word ``involve'' with 
the word ``require'' as it applies to changes made to the updated Final 
Safety Analysis Report and the TS Bases.
    Date of issuance: July 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 200 & 193.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10010). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 20, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 5.5.14 to eliminate the use of the term 
``unreviewed safety question,'' and replace the word ``involve'' with 
the word ``require'' as it applies to changes made to the updated Final 
Safety Analysis Report and the TS Bases.
    Date of issuance: July 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 204 & 185.
    Facility Operating License Nos. NPF-and NPF-17: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2921). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 17, 2002.
    No significant hazards consideration comments received: No.

Energy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Energy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: January 31, 2002, as 
supplemented by letter dated June 20, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification 3.8.1, ``AC Sources-Operating,'' to extend the allowed 
outage time for a Division 1 or Division 2 Diesel Generator from the 
current 72 hours to 14 days.
    Date of issuance: July 16, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 151.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15623). The June 20, 2002, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 16, 2002.
    No significant hazardous consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-a254 and 50-265, Quad 
Cites Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: May 1, 2002.
    Brief description of amendments: The amendments revise the start 
delay time in the surveillance for the emergency diesel generators from 
``[le]10 seconds'' to ``[le]13 seconds.''
    Date of issuance: July 17, 2002.
    Effective date: For Unit 2, as of the date of issuance and shall be 
implemented within 30 days of the completion of Unit 1 refueling outage 
17, which is scheduled for November 2002. For Unit 1, as of the date of 
issuance and shall be implemented within 30 days following the date 
when General Electric (GE)-14 fuel is loaded into the reactor, which is 
scheduled during refueling outage 17 in November 2002. The amendment 
may not be implemented prior to the date GE-14 fuel is loaded into the 
reactor.
    Amendment Nos.: 206 and 202.

[[Page 50964]]

    Facility Operating License Nos. DPR-29 and DRP-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: a May 28, 2002 (67 FR 
36931). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 17, 2002.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendments: July 24, 2001, as supplemented 
June 5, and July 1.
    Brief description of amendments: The amendments revise the Improved 
Technical Specifications (ITS) to accommodate future changes in plant 
design, including increased levels of Once--Through Steam Generator 
(OTSG) tube plugging. The changes are categorized into two sets. The 
first set of changes relocate parameters from the ITS to the cycle-
specific Core Operating Limits Report (COLR). These parameters are the 
Variable Low Pressure Trip equation specified in ITS Table 3.3.1-1, and 
Reactor Coolant System (RCS) pressure limit within Surveillance 
Requirement (SR) 3.4.1.1. The second set of changes are applicable to 
raising the OTSG tube plugging limit to a maximum of 20% equivalent of 
all tubes, and addresses its impact. These changes include the revision 
of the hot leg maximum temperature limit, and the revision of the RCS 
minimum flow limits for four- and three-reactor coolant pump operation. 
The RCS limits associated with 20% tube plugging will be maintained in 
its ITS. Cycle-specific values of these limits, however, have been 
relocated to the COLR. The hot leg temperature and RCS flow limit 
values within SR 3.4.1.2 and 3.4.1.3 ``RCS Pressure, Temperature, and 
Flow DNB [departure from nucleate boiling] Limits,'' were relocated to 
reflect their location in the COLR. For both sets of changes, ITS 
5.6.2.18(a) was modified to reflect the relocation of cycle-specific 
values from the ITS and the COLR.
    Date of issuance: July 16, 2002.
    Effective date: As of the date of issuance shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 204.
    Facility Operating License Nos. DPR-72: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: a August 22, 2001 (66 
FR 44173). The supplemental letters provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 16, 2002.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: July 30, 2001, as supplemented by letter 
dated August 23, 2001.
    Description of amendment request: The amendment revises the Cooper 
Nuclear Station's licensing basis.
    Date of issuance: July 19, 2002.
    Effective date: The amendment is effective on the date of issuance, 
to be implemented within 30 days from the date of issuance.
    Amendment No.: 192.
    Facility Operating License No. DPR-46: Amendment revises the Cooper 
Nuclear Station's licensing basis.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. June 25, 2002 (67 FR 42828). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided an opportunity to request a hearing by July 29, 2002, but 
indicated that, if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated July 19, 2002.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station (CNS), Nemaha County, Nebraska.

    Date of amendment request: May 20, 2002, as supplemented by letters 
dated June 19, July 3 (two letters), and July 12, 2002. The letters 
dated July 3 (two letters), and July 12, 2002, were of a clarifying 
nature, did not expand the application beyond the scope of the initial 
notice, and did not affect the staff's initial proposed no significant 
hazards consideration determination.
    Description of amendment request: The amendment revises the Cooper 
Nuclear Station's Technical Specifications (TS) 3.7.2 and 3.7.3 
reflecting increases in TS temperature limits for ultimate heat sink 
and reactor equipment cooling water temperatures.
    Date of issuance: July 22, 2002.
    Effective date: The amendment is effective on the date of issuance, 
to be implemented within 30 days from the date of issuance.
    Amendment No.: 193.
    Facility Operating License No. DPR-46: Amendment revises the Cooper 
Nuclear Station's TS.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. 67 FR 43688 dated June 28, 2002. The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided an opportunity to request a hearing by July 12, 2002, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated July 22, 2002.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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 application for amendments: September 20, 2001, as 
supplemented by letters dated March 27 and April 12, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications to support extension of the operating cycle 
from 18 months to 24 months.
    Date of issuance: July 12, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 232/174.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59512). The supplements dated March 27 and April 12, 2002, provided 
clarifying information that did not

[[Page 50965]]

change the scope of the September 20, 2001, application nor the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 12, 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: October 22, 2001, as supplemented by 
letters dated May 16 and June 25, 2002.
    Brief description of amendments: The amendments change TS 3/4.9.4, 
``Refueling Operations--Containment Building Penetrations'', to allow 
the equipment hatch to be open during core alterations or movement of 
irradiated fuel within the containment.
    Date of issuance: July 18, 2002.
    Effective date: July 18, 2002.
    Amendment Nos.: Unit 1--139; Unit 2--128.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2930). The May 16 and June 25, 2002, supplemental letters provided 
clarifying information that was within the scope of the original 
Federal Register notice and did not change the staff's initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 18, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 26th day of July 2002.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 02-19420 Filed 8-5-02; 8:45 am]
BILLING CODE 7590-01-P