[Federal Register Volume 68, Number 52 (Tuesday, March 18, 2003)]
[Notices]
[Pages 12946-12964]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-6286]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, February 21, 2003, through March 6, 2003. 
The last biweekly notice was published on March 4, 2003 (68 FR 10277).

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

[[Page 12947]]

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 Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By April 17, 2003, 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, 
Public File Area 01F21, 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.
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    \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 paragraphs (d)(1) and (d)(2) 
regarding petitions to intervene and contentions. For the complete, 
corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April 
29, 2002.
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    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, Public File Area 01F21, 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 12948]]

    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, Public File 
Area 01F21, 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, 301-415-4737 or by e-mail to 
[email protected].

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

    Date of amendment request: January 16, 2003
    Description of amendment request: The proposed amendment would 
revise the TMI-1 Technical Specifications to incorporate changes 
associated with the Cycle 15 core reload design analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specification limits (Figure 2.1-1) and 
reactor protection system (RPS) trip setpoints (Table 2.3-1) are 
developed in accordance with the methods and assumptions described 
in NRC-[Nuclear Regulatory Commission] approved Framatome ANP 
Topical Reports BAW-10179 P-A, ``Safety Criteria and Methodology for 
Acceptable Cycle Reload Analyses'' and BAW-10187 P-A, ``Statistical 
Core Design for B&W-[Babcock&Wilcox-] Designed 177 FA Plants.'' The 
core thermal-hydraulic code (LYNXT) and CHF [critical heat flux] 
correlation (BWC) have been approved for use with these methods and 
the Mark-B fuel type utilized at TMI Unit 1. The proposed Technical 
Specification requirements on Variable Low Pressure Trip (VLPT) 
instrument operating conditions (Table 3.5-1) and surveillances 
(Table 4.1-1) are consistent with the VLPT requirements that were 
last contained in the TMI Unit 1 Technical Specifications prior to 
Cycle 7. The existing flux-flow trip setpoint and power/pump monitor 
trip have been shown to provide adequate DNB [departure from 
nucleate boiling] protection for Updated Final Safety Analysis 
Report (UFSAR) DNB-limiting loss of coolant events.
    The margin retained for penalties such as transition core 
effects, by imposing a Thermal Design Limit of 1.40 in all DNB 
analyses supporting the proposed change, has been shown to be 
sufficient to offset the current mixed core conditions at TMI Unit 
1, where the Mark-B12 fuel design with fine mesh debris filter is 
co-resident with earlier, non-debris filter Mark-B fuel designs. 
Therefore the previous commitment to require a higher minimum RCS 
[reactor coolant system] flow (105.5% of design flow instead of 
104.5%) to offset transition core penalties is no longer necessary.
    Reload cycles are designed and operated with maximum steady-
state radial-local peaking factors that are bounded by UFSAR 
assumptions used to determine the dose consequences from fuel 
handling accidents.
    The proposed change to Technical Specification 3.5.2.2.a is only 
an administrative correction.
    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?
    Response: No.
    The proposed Technical Specification limits (Figure 2.1-1) and 
reactor protection system (RPS) trip setpoints (Table 2.3-1) provide 
core protection safety limits and Variable Low Pressure Trip 
setpoints developed in accordance with NRC-approved methods and 
assumptions. The transition core penalty resulting from Mark-B12 
fuel with fine mesh debris filters co-residing with earlier, non 
debris filter Mark-B fuel has been demonstrated to be sufficiently 
bounded by the analyses supporting the proposed amendment. Therefore 
the previous commitment to require a higher minimum RCS flow (105.5% 
of design flow instead of 104.5%) to offset transition core 
penalties is no longer necessary. These changes have been evaluated 
for their impact on the design and operation of plant structures, 
systems, and components. These changes do not introduce any new 
accident precursors and do not involve any alterations to plant 
configurations, which could initiate a new or different kind of 
accident.
    The proposed change to Technical Specification 3.5.2.2.a is only 
an administrative correction.
    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 a 
margin of safety?
    Response: No.
    The proposed reactor protection system (RPS) trip setpoints 
(Table 2.3-1) ensure core protection safety limits will be preserved 
during power operation. The proposed safety limits and setpoints are 
developed in accordance with NRC-approved methods and assumptions. 
The margin retained for penalties such as transition core effects, 
by imposing a Thermal Design Limit of 1.40 in all DNB analyses 
supporting the proposed change, has been shown to be sufficient to 
offset the current mixed core conditions at TMI Unit 1. The margin 
available between minimum DNBR [departure from nucleate boiling 
ratio] results for UFSAR loss of coolant flow events and the Thermal 
Design Limit of 1.40 is significant and is similar to DNB margin 
results for the current non-SCD [Statistical Core Design] analysis.
    Reload cycles are designed and operated with maximum steady-
state radial-local peaking factors that are bounded by UFSAR 
assumptions used to determine the dose consequences from fuel 
handling accidents.
    The proposed change to Technical Specification 3.5.2.2.a is only 
an administrative correction.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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: February 14, 2003.

[[Page 12949]]

    Description of amendment request: The amendment would allow an 
increase in the maximum decay heat of spent fuel stored in Spent Fuel 
Pools (SFPs) C and D from 1.0 MBTU/hr to 7.0 MBTU/hr in Technical 
Specification 5.6.3.d. The amendment would also increase the allowable 
SFP temperatures from 140 degrees F to 150 degrees F under normal and 
emergency conditions other than a design-basis Loss-of-Coolant Accident 
(LOCA). For a LOCA, the maximum allowed SFP temperature would increase 
from 150 degrees F to 160 degrees F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A written evaluation of the significant hazards consideration of 
a proposed license amendment is required by 10 CFR 50.92. Progress 
Energy Carolinas, Inc. (alternately known as Carolina Power & Light 
Company) has evaluated the proposed amendment and determined that it 
involves no significant hazards consideration. According to 10 CFR 
50.92, a proposed amendment to an operating license involves no 
significant hazards consideration if 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 determination is as follows:

Proposed Change

    The change involves an increase in the maximum decay heat of 
spent fuel stored in Spent Fuel Pools (SFPs) C and D from 1.0 MBTU/
hr to 7.0 MBTU/hr, and an increase in the allowable SFP 
temperatures.

Basis

    This change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The license amendment only increases the heat load from the Fuel 
Pool Cooling and Cleanup System (FPCCS) and the maximum allowable 
pool temperature. The changes do not modify the design of 
Structures, Systems and Components (SSCs) that could initiate an 
accident. The FHB [Fuel Handling Building] Emergency Exhaust System 
mitigates the consequences of a fuel handling accident in the Fuel 
Handling Building. This system has been evaluated for the conditions 
that would exist with the higher SFP temperatures and it was found 
that there would be no decrease in the charcoal efficiency. As a 
result, there was no increase in the doses from the fuel handling 
accident in the FHB. Therefore, the change does not result in any 
increase in the probability or consequences in any accident 
previously analyzed.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The increase in the SFP decay heat load and the SFP temperature 
limit does not involve new plant components or procedures. No 
significant impact on any postulated accident is made due to this 
change since the required cooling capacity is maintained to the SFPs 
and the FPCCS, and the SFPs will operate within design parameters.
    For the activation of SFPs C and D, Progress Energy Carolinas, 
Inc. performed a Probabilistic Safety Analysis (PSA) of a total loss 
of SFP forced cooling. That analysis concluded that the probability 
of spent fuel rack uncovery was not credible. That analysis remains 
bounding for this license amendment application.
    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 
the barriers to fission product release (fuel cladding, reactor 
coolant system pressure boundary, and containment boundary). The 
change in the SFPs C and D decay heat load is bounded by the heat 
load used in the analysis of the safety-related systems for design 
basis accidents. Therefore, there is no impact in the margin of 
safety.
    Based on these considerations, the proposed change does not 
involve a significant reduction on 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: Allen Howe.

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: February 19, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 5.5.10, ``Steam Generator (SGs) 
Tube Surveillance Program.'' The proposed amendments would relocate to 
TS 5.5.21 the TS 5.5.10 program requirements that apply to the original 
SGs and would provide a new TS 5.5.10 that contains program 
requirements that would apply to the new SGs when they are installed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.91, Duke has made the determination that 
this amendment request does not involves a significant hazard by 
applying the three standards established by the NRC regulations in 
10 CFR 50.92 as described below.

First Standard

    The proposed amendment would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment will revise Technical Specification (TS) 
5.5.10 to delete and clarify replacement steam generator (SG) 
surveillance requirements applicable to the replacement of the SGs 
following their installation. The proposed amendment does not result 
in any changes to the design or methods of operation of the facility 
or any of its structures, systems or components (SSC). The SG repair 
methods that would be deleted are not applicable to the replacement 
SGs due to the use of improved materials and design. Defects found 
during future replacement SG tube inspections that exceed the limits 
in the new TS 5.5.10 will be removed from service by plugging rather 
than being repaired. The accident analyses and assumptions made in 
the Updated Final Safety Analysis Report (UFSAR) Chapter 15, 
Accident Analyses, are not changed as a result of the proposed 
changes. There are no changes resulting from the new TS 5.5.10 that 
could affect the function of preventing or mitigating any of these 
accidents. The proposed change does not increase the likelihood of 
the malfunction of an SSC that may increase the probability or 
consequences of an accident. The relocated surveillance requirements 
for the current steam generators will not change as a result of the 
proposed TS changes. Therefore, the proposed change will not result 
in a significant increase in the probability or consequences of an 
accident previously evaluated.

Second Standard

    The proposed amendment would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the SG tube surveillance TS will delete 
or modify surveillance requirements that would otherwise not be 
applicable to the replacement steam generators. SG Tubes found to 
exceed the plugging limit criteria of TS 5.5.10 for continued

[[Page 12950]]

service will be removed from service by plugging rather than being 
repaired. The plugging limit is unchanged by the proposed amendment. 
These changes will not introduce any adverse changes to the 
facilities' design bases or postulated accidents resulting from 
potential tube degradation. The proposed amendment does not affect 
the design of SGs, their method of operation, or primary coolant 
chemistry controls. In addition, the proposed amendment does not 
impact any other SSC. Surveillance requirements for the current SGs 
will not change prior to their removal from service as a result of 
the proposed changes. Therefore, the proposed changes do not create 
the possibility of a new or different type of accident from any 
accident previously evaluated.

Third Standard

    The proposed amendment would not involve a significant reduction 
in the margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. These barriers are unaffected by the changes proposed in 
this LAR. The steam generator tubes are an integral part of the 
reactor coolant pressure boundary. Repairing SG tubes by previously 
approved methods of sleeving or rerolling are considered to be an 
equivalent boundary to plugging a steam generator tube as has also 
been previously approved. Therefore, the margin of safety is not 
reduced by the changes proposed in this license amendment request.

Conclusion

    Based upon the proceeding evaluation, performed pursuant to 10 
CFR 50.92, Duke Energy Corporation has concluded that approval and 
implementation of this license amendment request at the Oconee 
Nuclear Station will not involve a significant hazards 
consideration. The proposed changes revise the steam generator 
surveillance requirements to be consistent with the replacement 
steam generators. Following implementation of the changes proposed 
in this license amendment request, the Oconee steam generators will 
continue to be operated in a safe and conservative manner.

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

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

    Date of amendment request: January 29, 2003.
    Description of amendment request: The proposed amendment would 
change the spent fuel pool loading restrictions by redefining the 
regions, inserting Metamic[reg] poison panels in a portion of the spent 
fuel pool, and increasing the minimum boron concentration.
    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.
    Most accident conditions will not result in an increase in K-
effective (Keff) of the fuel stored in the rack. However, 
there are accidents that can be postulated to increase reactivity. 
For these accident conditions, the double contingency principle of 
ANS [American Nuclear Society] N16.1-1975 is applied. This states 
that it is unnecessary to assume two unlikely, independent, 
concurrent events to ensure protection against a criticality 
accident. Therefore, for accident conditions, the presence of 
soluble boron in the storage pool water can be assumed as a 
realistic initial condition since its absence would be a second 
unlikely event.
    A vertical drop accident condition directly upon a cell will 
cause damage to the racks in the active fuel region. The proposed 
2000 ppm [parts per million] TS [technical specification] 
limit will insure that Keff does not exceed 0.95. A fuel 
assembly dropped on top of the rack will not deform the rack 
structure such that criticality assumptions are invalidated. The 
rack structure is such that [after rack deformation] an assembly 
positioned horizontally on top of the rack is more than eight inches 
away from the upper end of the active fuel region of the stored 
assemblies. This distance precludes interaction between the dropped 
assembly and the stored fuel. An inadvertent drop of an assembly 
between the outside periphery of the rack and the pool wall is 
bounded by the worst case fuel misplacement accident condition of 
825 ppm. The distance between all the rack modules and the pool 
walls is [nominally] less than the width of a fuel assembly.
    The fuel assembly misplacement accident was considered for all 
storage configurations. An assembly with high reactivity is assumed 
to be placed in a storage location which requires restricted storage 
based on initial U-235 [Uranium-235] loading and burnup. The 
presence of boron in the pool water assumed in the analysis has been 
shown to substantially offset the worst case reactivity effect of a 
misplaced fuel assembly for any configuration. The boron requirement 
of 825 ppm is less than the proposed 2000 ppm minimum 
boron TS limit. Therefore, a five percent subcriticality margin can 
be easily met for postulated accidents since any reactivity increase 
will be much less than the negative worth of the dissolved boron.
    For fuel storage applications, water is present. An ``optimum 
moderation'' accident is not a concern in spent fuel pool storage 
racks because the rack design prevents the preferential reduction of 
water density between the cells of a rack (e.g., boiling between 
cells).
    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?
    Response: No.
    The proposed changes will define a portion of the current Region 
2 as Region 3. The new region will contain Metamic[reg] poison panel 
inserts and will allow unrestricted storage of fuel assemblies with 
various enrichments and burnup. To support the proposed change, a 
new criticality analysis was performed. The analysis resulted in new 
loading restrictions in Region 1 and Region 2. The presence of boron 
in the pool water assumed in the analysis is less than the proposed 
ANO-2 [Arkansas Nuclear One, Unit 2] TS minimum concentration of 
2000 ppm. Therefore, a five percent subcriticality margin 
can be easily met for postulated accidents since any reactivity 
increase will be much less than the negative worth of the dissolved 
boron.
    No new or different types of fuel assembly drop scenarios are 
created by the proposed change. During the installation of the 
Metamic[reg] panels, the possible drop of a panel is bounded by the 
current fuel assembly drop analysis. No new or different fuel 
assembly misplacement accidents will be created. Administrative 
controls currently exist to assist in assuring that fuel 
misplacement does not occur.
    Therefore, the proposed change does 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?
    Response: No.
    With the presence of a nominal boron concentration, the SFP 
[spent fuel pool] storage racks are designed to assure that fuel 
assemblies of less than or equal to five weight percent U-235 
enrichment when loaded in accordance with the proposed loading 
restrictions will be maintained within a subcritical array with a 
subcritical margin of

[[Page 12951]]

five percent. This has been verified by criticality analyses.
    Credit for soluble boron in the SFP water is permitted under 
accident conditions. The proposed change that will allow insertion 
of Metamic[reg] poison panels does not result in the potential of 
any new misplacement scenarios. Criticality analyses have been 
performed to determine the required boron concentration that would 
ensure that the maximum Keff does not exceed 0.95. By 
increasing the minimum boron concentration to 2000 ppm, 
the margin of safety currently defined by taking credit for soluble 
boron will be maintained.
    The structural analysis of the spent fuel racks along with the 
evaluation of the SFP structure showed that the integrity of these 
structures will be maintained with the addition of the poison 
inserts. All structural requirements were shown to be satisfied, so 
all the safety margins were maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: December 17, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.10, ``Ventilation Filter 
Testing Program,'' to adopt the requirements of the American Society 
for Testing and Materials Standard (ASTM) D3803-1989, ``Standard Test 
Method for Nuclear-Grade Activated Carbon.'' The proposed TS revisions 
are in response to Nuclear Regulatory Commission (NRC) Generic Letter 
(GL) 99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' 
The NRC had previously published a notice of consideration on December 
12, 2001 (66 FR 64292) regarding a similar proposal from the licensee 
in response to GL 99-02. However, in response to a request for 
additional information from the NRC dated March 29, 2002, the licensee 
has now revised its proposed amendment. In addition to withdrawing the 
prior request to change the maximum control room ventilation system 
(CRVS) differential pressure in TS 5.5.10.d, the proposed amendment 
would revise the TSs: (1) To provide a CRVS methyl iodide removal 
efficiency of greater than or equal to 95.5% and remove the notation 
that there is a 1-inch charcoal bed depth; (2) to allow for the 
continued use of the existing CRVS through Refueling Outage 13, in 
order to design, fabricate, and install a 2-inch charcoal filter bed; 
(3) to add a note in the TS requiring a demonstration of charcoal 
efficiency of 93% when changing the charcoal in the existing CRVS bed 
prior to any fuel movement in the upcoming Refueling Outage 12 and 
every 6 months thereafter until the new beds are installed. The 
proposed amendment also seeks an exception from the factor of safety of 
two for the Containment Fan Cooler Units due to the plant's design.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: The proposed license amendment adopts the new test 
method and acceptance criteria of ASTM D3803-1989 for activated 
charcoal filters. The changes require laboratory performance testing 
of adsorber carbon that yields a more accurate result than the 
testing currently required by the TS. The proposed change to delete 
non-conservative TS requirements for testing of adsorber carbon is 
not a plant accident initiator as described in the Final Safety 
Analysis Report (FSAR). The proposed amendment does not change the 
function of any structure, system or component (SSC). The function 
of the ventilation systems is filtration of radiological releases 
during postulated accidents. The proposed changes will provide 
greater assurance that this function is provided. The revised TS 
requirements are for laboratory tests that are currently in place to 
address Generic Letter 99-02, with one exception to the safety 
factor of 2, and accommodate the change of the Control Room 
Ventilation System (CRVS) charcoal beds to two inches. The change 
only affects the TS testing requirements since the modification to 
the CRVS will be accomplished separately from the TS change. The TS 
changes will not result in any changes to the efficiency assumed in 
accident analysis. The changes do not alter, degrade or prevent 
actions described or assumed in an accident described in the FSAR. 
Therefore, the proposed amendment does not change the possibility of 
an accident previously evaluated or significantly increase the 
consequences of an accident previously evaluated.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed license amendment adopts the new test 
method and acceptance criteria of ASTM D3803-1989 for activated 
charcoal filters. The change does not involve any modifications to 
the plant but will accommodate the planned modification of the CRVS 
to change the charcoal beds from 1 inch to 2 inches. The change will 
not require changes to how the plant is operated nor will it affect 
the operation of the plant. The changes require laboratory 
performance testing of adsorber carbon that yields a more accurate 
result than the testing currently required by the TS. The proposed 
changes to delete non-conservative TS requirements for testing of 
adsorber carbon is not a plant accident initiator as described in 
the Final Safety Analysis Report (FSAR). The proposed amendment does 
not change the function of any structure, system or component (SSC). 
The function of the ventilation systems is filtration of 
radiological releases during postulated accidents. The proposed 
changes will provide greater assurance that this function is 
provided. The revised TS requirements are for laboratory tests that 
are currently in place to address Generic Letter 99-02, with one 
exception to the safety factor of 2, and accommodate the change of 
the Control Room Ventilation System (CRVS) charcoal beds to two 
inches. The change only affects the TS testing requirements since 
the modification to the CRVS will be accomplished separately from 
the TS change. The TS changes will not result in any changes to the 
efficiency assumed in accident analysis. The changes do not alter, 
degrade or prevent actions described or assumed in an accident 
described in the FSAR. Therefore, the proposed amendment does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    Response: The proposed license amendment adopts the new test 
method and acceptance criteria of ASTM D3803-1989 for activated 
charcoal filters. The proposed license amendment does not reduce the 
margin of safety but enhances it by requiring more accurate testing. 
The proposed test change will require the use of a current and 
improved ASTM standard to ensure that the carbon ability to adsorb 
radioactive material will remain at or above the capability credited 
in our accident analysis.

    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.

[[Page 12952]]

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: January 23, 2003.
    Description of amendment request: The proposed amendment would 
modify the Pilgrim Nuclear Power Station Technical Specification (TS) 
requirements for the Emergency Core Cooling System (ECCS) during 
shutdown conditions. The proposed amendment would change the Core Spray 
and Low Pressure Coolant Injection System's TS requirements to be 
applicable during the Run, Startup, and Hot Shutdown Modes. The 
proposed change would also modify the High Drywell Pressure 
Instrumentation TSs to require the instrumentation to be Operable 
during the Run, Startup and Hot Shutdown Modes. The proposed change 
would also remove unnecessary TS requirements based on the plant's 
operating Mode. Other proposed changes are administrative in nature.
    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 review is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change involves modifications to the TS operability 
requirements for the ECCS during shutdown conditions. The ECCS is 
designed to mitigate the release of radioactive materials to the 
environment following a Loss of Coolant Accident (LOCA). The 
modifications remove certain ECCS TS requirements during shutdown 
conditions and includes additional requirements for the Cold 
Shutdown or Refuel Modes when the availability of the ECCS is most 
likely to be needed. The additional requirements are more 
restrictive and are proposed to reduce the probability or 
consequences of potential accidents. The requirements proposed to be 
removed are unnecessary due to the associated plant conditions and 
other changes are administrative in nature. No increase in the 
probability or consequences of an accident previously evaluated has 
been identified for these changes. The ECCS is not an initiator of 
any accidents previously evaluated and the proposed change does not 
increase the amount of radioactive materials available to be 
released for a previously evaluated accident. 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?
    The proposed change involves modifications to the TS operability 
requirements for the ECCS during shutdown conditions. The 
modifications remove unnecessary ECCS TS requirements during 
shutdown conditions and includes additional requirements for the 
Cold Shutdown or Refuel Modes when the availability of the ECCS is 
most likely to be needed. In addition, the proposed change makes 
administrative changes. The proposed change does not involve any 
physical alteration of ECCS equipment and does not create a new mode 
of system operation. In addition, no new or different types of ECCS 
equipment will be installed as a result of the proposed change. The 
proposed change will allow the installation of modifications on the 
reference and variable legs of the instrument racks that support the 
ECCS and Feedwater level instrumentation. No other types of 
accidents or accident initiators associated with the proposed change 
or modifications have been identified. 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 a 
margin of safety?
    The ECCS is designed to mitigate the release of radioactive 
materials to the environment following a LOCA. The long-term cooling 
analysis following a design basis LOCA demonstrates that only one 
low-pressure ECCS injection/spray subsystem is required, post LOCA, 
to maintain adequate reactor vessel water level. The proposed change 
includes an additional requirement that two low-pressure injection/
spray subsystems be Operable for the Cold Shutdown or Refuel Modes. 
The requirements proposed to be removed are unnecessary due to the 
associated plant conditions and other proposed changes are 
administrative in nature. No scenario has been identified that, as a 
result of the proposed change, would create a single component 
failure which prevents the automatic initiation of the ECCS. The 
proposed change will not modify the method by which any safety-
related system performs its function and ECCS operation and testing 
will remain consistent with current safety analysis assumptions. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: December 20, 2002.
    Description of amendment request: The proposed amendments would 
remove technical specification requirements for reactor protection 
system Function 5, main steam isolation valve closure, and Function 10, 
turbine condenser vacuum low, when in startup.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to the Dresden Nuclear Power Station (DNPS) 
Units 2 and 3 Technical Specifications (TS) revise the applicability 
of TS 3.3.1.1, ``Reactor Protection System (RPS) Instrumentation,'' 
Function 5 (i.e., Main Steam Isolation Valve--Closure) and Function 
10 (i.e., Turbine Condenser Vacuum--Low) to eliminate the 
requirement for these functions to be operable while in Mode 2 with 
reactor pressure =600 psig. The proposed changes also 
delete Required Action F.2 of TS 3.3.1.1 to align with the revised 
applicability for Functions 5 and 10.
    TS requirements that govern operability or routine testing of 
plant instruments are not assumed to be initiators of any analyzed 
event because these instruments are intended to prevent, detect, or 
mitigate accidents. Therefore, these proposed changes will not 
involve an increase in the probability of an accident previously 
evaluated.
    Additionally, these proposed changes will not increase the 
consequences of an accident previously evaluated because the 
proposed changes do not adversely impact structures, systems, or 
components. These changes will not alter the operation of equipment 
assumed to be available for the mitigation of accidents or 
transients by the plant safety analysis. Functions 5 and 10 are 
currently required in Mode 2 with reactor pressure =600 
psig to ensure that the reactor is shut down to prevent an 
overpressurization transient due to closure of main steam isolation 
valves or turbine stop valves. The existing scram logic is the 
result of experience gained during the startup of an early vintage 
boiling water reactor in 1966 when operators had difficulty 
controlling reactor power above approximately 600 psig without 
pressure control. Experience on later plant startups indicates that 
the early experience may not be inherent to the boiling water 
reactor design. As such, General Electric subsequently recommended 
that the scram requirement be eliminated. In Mode 2, the heat 
generation rate is low enough so that the

[[Page 12953]]

other diverse RPS functions provide sufficient protection from an 
overpressurization transient. Furthermore, there will be no change 
in the types or significant increase in the amounts of any effluents 
released offsite.
    For these reasons, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes revise the applicability for Functions 5 
and 10 of TS 3.3.1.1. The RPS is not an initiator of any accident. 
Rather, the RPS is designed to initiate a reactor scram when one or 
more monitored parameters exceed their specified limits to preserve 
the integrity of the fuel cladding and the reactor coolant pressure 
boundary and minimize the energy that must be absorbed following an 
accident. The proposed changes do not alter the applicability for 
RPS functions during plant conditions in which an overpressurization 
transient is assumed to occur. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
and actions. The proposed changes revise the applicability for 
Functions 5 and 10 of TS 3.3.1.1. The proposed changes do not alter 
the applicability for RPS functions during plant conditions in which 
an overpressurization transient is assumed to occur. In addition, 
the proposed changes do not affect the probability of failure or 
availability of the affected instrumentation. Furthermore, the 
proposed changes will reduce the probability of test-induced plant 
transients and equipment failures. 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: February 4, 2003.
    Description of amendment request: The proposed amendment would 
extend the surveillance interval of the slave relay in the Engineered 
Safety Feature Actuation System instrumentation from 92 days to 12 
months. The proposed amendment includes changes to surveillance 
requirement (SR) 4.3.2.1.1 and the related Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The proposed change to the slave relay test 
interval reduces the potential for spurious actuation of equipment, 
and therefore does not increase the probability of any accident 
previously analyzed. The proposed change to the slave relay test 
interval does not change the response of the unit to any accidents 
and has an insignificant impact on the reliability of the engineered 
safety feature actuation system (ESFAS) signals. The ESFAS will 
remain highly reliable and the proposed change will not result in a 
significant increase in the risk of plant operation. This is 
demonstrated by showing that the impact on plant safety as measured 
by the change in core damage frequency (CDF) is less than 1.0E-06 
per year and the change in large early release frequency (LERF) is 
less than 1.0E-07 per year. The change meets the acceptance criteria 
in Regulatory Guide 1.174. Therefore, since the ESFAS will continue 
to perform its function with high reliability as originally assumed, 
and the increase in risk as measured by the change in CDF and LERF 
is within the acceptance criteria of existing regulatory guidance, 
there will not be a significant increase in the consequences of any 
accidents.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the unit is operated and maintained. The proposed change does not 
alter or prevent the ability of structures, systems, and components 
(SSCs) from performing their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Further, the proposed change does not increase the types 
or amounts of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed change is consistent with the 
safety analysis assumptions and resultant consequences.
    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?
    Response: No. The proposed change does not result in a change in 
the manner in which the EFSAS provides unit protection. The EFSAS 
will continue to have the same setpoints after the proposed change 
is implemented. There are no design changes associated with the 
proposed change. The change to the slave relay test interval does 
not change any existing accident scenarios, nor create any new or 
different accident scenarios.
    The change does not involve a physical alteration to the unit 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the change does not impose any new or different requirements or 
eliminate any existing requirements. The change does not alter 
assumptions made in the safety analysis. The proposed change is 
consistent with the safety analysis assumptions and current unit 
operating practice.
    Therefore, the proposed change does 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?
    Response: No. The proposed change does not alter the manner in 
which safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not impacted by this change. Redundant ESFAS 
trains are maintained, and diversity with regard to the signals that 
provide engineered safety features actuation is also maintained. All 
signals credited as primary or secondary, and all operator actions 
credited in the accident analysis will remain the same. The proposed 
change will not result in unit operation in a configuration outside 
the design basis. The calculated impact on risk is insignificant and 
meets the acceptance criteria contained in Regulatory Guide 1.174. 
The proposed slave relay test interval change will result in a 
reduced potential for spurious equipment actuations associated with 
testing.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 12954]]

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: June 10, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated June 10, 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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

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

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

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

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

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

    Date of amendment request: February 28, 2003.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to relocate the numerical 
values and curves for the pressure and temperature (P/T) limits for the 
reactor coolant system (RCS). The numerical values and curves would be 
relocated from the TS to a licensee-controlled document, the Pressure 
and Temperature Limits Report (PTLR) pursuant to Nuclear Regulatory 
Commission (NRC) Generic Letter (GL) 96-03, ``Relocation of the 
Pressure Temperature Limit Curves and Low Temperature Overpressure 
Protection System Limits,'' dated January 31, 1996, as modified by NRC 
Improved Standard TS, TS Task Force (TSTF) change package number 419, 
Revision 0. Specifically, a definition for the PTLR would be added to 
TS 1.0, ``Definitions;'' administrative controls for the generation and 
reporting requirements associated with the PTLR would be added to TS 
5.6, ``Administrative Controls--Reporting Requirements; '' TSs 3.4.9 
and 4.4.9 would be modified by removing the numerical values and curve 
(Figure 3.4.9-1) for the various P/T limits (which the licensee has 
updated using an NRC-approved methodology) and replacing them with a 
reference to the PTLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The P/T limits are not derived from Design Basis Accident (DBA) 
analyses. They are prescribed by the ASME [American Society of 
Mechanical Engineers Boiler and Pressure Vessel] Code and 10 CFR 
[Part] 50 Appendi[ces] G and H as restrictions on normal operation 
to avoid encountering

[[Page 12955]]

pressure, temperature, and temperature rate of change conditions 
that might cause undetected flaws to propagate and cause non-ductile 
failure of the reactor coolant pressure boundary. Thus, they ensure 
that an accident precursor is not likely. Hence, they are included 
in the TS as satisfying Criterion 2 of 10 CFR 50.36(c)(2)(ii). The 
relocation of the numerical value of these limits to a licensee-
controlled document does not remove the existing TS requirement that 
the limits be met. The new TS administrative controls for the PTLR 
will ensure that only NRC-approved methods are used to calculate the 
actual limits to be applied. Thus, this relocation will not increase 
the probability of any accident previously evaluated.
    The proposed changes do not alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the facility is operated or maintained. The proposed changes will 
not affect any other System, Structure or Component (SSC) designed 
for the mitigation of previously analyzed events. The proposed 
changes do not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of any accident previously evaluated. Thus, the 
proposed relocation of the existing numerical values and the updated 
figure for the RCS P/T limits based upon an NRC-approved 
methodology, to a licensee-controlled document (i.e., the PTLR), 
with all the requisite TS restrictions placed upon it by NRC Generic 
Letter 96-03, as modified by TSTF-419, Rev. 0, will not increase the 
consequences of any previously evaluated accident.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. We are merely requesting to 
move the existing numerical values and the updated figure for the 
RCS P/T limits based upon an NRC-approved methodology, from the TS 
to a licensee-controlled document (i.e., the PTLR), with all the 
requisite TS restrictions placed upon it by NRC Generic Letter 96-
03, as modified by TSTF-419, Rev. 0.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed changes do not alter the manner in which Safety 
Limits, Limiting Safety System Settings or Limiting Conditions for 
Operation are determined. The setpoints at which protective actions 
are initiated are not altered by the proposed changes. Sufficient 
equipment remains available to actuate upon demand for the purpose 
of mitigating an analyzed event. We are merely requesting to move 
the existing numerical values and the updated figure for the RCS P/T 
limits based upon an NRC-approved methodology, from the TS to a 
licensee-controlled document (i.e., the PTLR), with all the 
requisite TS restrictions placed upon it by NRC Generic Letter 96-
03, as modified by TSTF-419, Rev. 0. Thus, the proposed changes will 
not significantly reduce any margin of safety that currently exists.
    Based upon the above, NMC [Nuclear Management Company] has 
determined that the proposed amendment will not involve a 
significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111 
Pennsylvania Avenue NW Washington, DC 20004.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: January 27, 2003.
    Description of amendment request: The proposed amendment would make 
administrative and editorial changes to the Fort Calhoun Station (FCS) 
Technical Specifications (TS) 1.3 Basis (1); 2.7(1)a; 2.7(1)b; 2.7(1)d; 
2.7(1)i; 2.7 Basis; 3.0.2; Table 3-5, Item 11; and 3.5(3)ii. The 
proposed changes consist primarily of editorial and typographical 
changes or corrections.
    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 correction of typographical errors and clarification of 
specifications is not an initiator of any previously evaluated 
accident. The frequency or periodicity of performance of those 
surveillances affected by this change are not an initiator of any 
previously evaluated accident. The proposed changes will not prevent 
safety systems from performing their accident mitigation function as 
assumed in the safety analysis.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change only affects the technical specifications 
and does not involve a physical change to the plant. Modifications 
will not be made to existing components nor will any new or 
different types of equipment be installed. The proposed change 
corrects typographical errors, provides clarification as to 
applicable equipment and modifies the frequency of surveillances 
performed once per shift from 8 hours to 12 hours. This change will 
not alter assumptions made in safety analysis and licensing bases.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change corrects typographical errors, provides 
clarification as to applicable equipment, and modifies the frequency 
of surveillances performed once per shift from 8 hours to 12 hours. 
The decrease in frequency or periodicity of performance of these 
surveillances will also permit more efficient and more safely 
managed plant operations and can help reduce the risk associated 
with changing plant equipment or operating modes in order to obtain 
some of these readings.
    Therefore, this technical specification 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: January 27, 2003.
    Description of amendment request: The proposed amendment would 
delete the allowance to perform the surveillance test of Table 3-2, 
Item 20 (Recirculation Actuation Logic Channel Functional Test) under 
administrative controls, while components in excess of those allowed by 
Conditions a, b, d, and e of Technical Specification 2.3(2) are 
inoperable provided they are returned to operable status within one 
hour. This allowance was granted in Amendment No. 206 issued April 19, 
2002, on an exigent basis and applies only for the remainder of the 
current cycle. Omaha

[[Page 12956]]

Public Power District committed to submit a permanent resolution to 
this allowance and this license amendment request constitutes this 
permanent resolution.
    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.
    Deleting the requirement to perform the quarterly surveillance 
test of Table 3-2, Item 20 (Recirculation Actuation Logic Channel 
Functional Test) under administrative controls is acceptable since 
the performance of the recirculation actuation logic channel 
functional test is not identified as the initiator of any analyzed 
event. The proposed change will still require that the surveillance 
test be performed and the required ECCS [emergency core cooling 
system] systems to be available. This change will not alter 
assumptions relative to the mitigation of an accident or transient 
event. The performance of this activity has no effect on any 
accident scenario. Therefore, the proposed change does not involve a 
significant increase in the 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.
    This change only removes a short term allowance to utilize 
administrative controls in the performance of the recirculation 
actuation logic channel functional test. These proposed changes do 
not involve a physical alteration of the plant (no new or different 
type of equipment will be installed) or change the methods governing 
plant operation. The proposed change does not involve any physical 
changes to plant systems, structures or components (SSCs) or the 
manner in which these SSCs are operated, maintained, modified or 
inspected. Therefore, these changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The minimum numbers of ECCS components required by the FCS [Fort 
Calhoun Station] accident analyses will remain available. The 
proposed change to delete the short term allowance to utilize 
administrative controls in the performance of the recirculation 
actuation logic channel functional test will not significantly 
impact the availability or reliability of the plant's systems or 
their ability to respond to plant transients and accidents. The 
performance of this activity has no effect on any accident scenario. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: January 27, 2003.
    Description of amendment request: The proposed amendment would 
authorize the revision of the Fort Calhoun Station, Unit No. 1 Updated 
Safety Analysis Report (USAR). Section 14.16 and Figures 14.16-1 
through 14.16-4 of the USAR will be revised to reflect the use of the 
GOTHIC, version 7.0, computer code and the results associated with the 
updated containment pressure analyses for a loss-of-coolant accident 
and main steam line break. In addition, GOTHIC will be used for the 
analysis of future plant upgrades associated with containment response 
and will be maintained consistent with other NRC-approved Omaha Public 
Power District methodologies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes will not increase the probability or 
consequence of any accident based on the following:
    The proposed changes to Section 14.16 of the Updated Safety 
Analysis Report (USAR) and replacements for Figures 14.16-1 through 
14.16-4 is required due to using GOTHIC, version 7.0 and the updated 
containment pressure analyses. Demonstrating that containment 
pressure is maintained less than the containment design pressure is 
required by Fort Calhoun Station (FCS) design basis. Additionally, 
the analyses credit all modes of heat transfer defined by Reference 
10.5. Therefore, the updated containment pressure analyses using 
GOTHIC, version 7.0 is in compliance with FCS design basis. Changes 
to the containment pressure analyses for either a loss-of-coolant 
accident or main steam line break will be controlled by 10 CFR 
50.59. Therefore, the probability or consequence of any accident is 
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 revision does not change any equipment required to 
mitigate the consequences of an accident. The continued use of the 
same USAR administrative controls prevents the possibility of a new 
or different kind of accident. Since the proposed changes do not 
involve the addition or modification of equipment nor alter the 
design of plant systems, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The changes proposed do not change how design 
basis accident events are postulated nor do the changes themselves 
initiate a new kind of accident or failure mode with a unique set of 
conditions (proposed administrative controls). Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of GOTHIC, version 7.0 is in compliance with FCS design 
basis. Additionally, GOTHIC has been benchmarked to the current 
analysis of record for a loss-of-coolant accident and main steam 
line break using the NRC approved computer code CONTRANS. These 
benchmark models demonstrate that GOTHIC provides similar results to 
CONTRANS. Future updates of the containment pressure analyses will 
be conducted under the 10 CFR 50.59 process. The analyses will 
credit all available modes of heat transfer defined by Reference 
10.5. Additionally, the main steam line break containment evaluation 
model considers the leakage past the broken steam generator main 
feed isolation valve of 2.45% of full power flow or approximately 
195 gpm. Therefore, the proposed changes do not involve a 
significant reduction to 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: January 27, 2003.
    Description of amendment request: The proposed amendment revises 
Technical Specifications (TS) 2.1.6, 3.2

[[Page 12957]]

(Table 3-5), and 5.9.1c. For TS 2.1.6(1), Omaha Public Power District 
(OPPD) has proposed to increase the ``as-found'' pressurizer safety 
valve (PSV) lift setting tolerance band of +/-1% to +1%/-3% to allow 
for normal setpoint variance for Modes 1 and 2. The Basis of TS 2.1.6 
will be revised to clarify that the PSVs are still operable and capable 
of performing their safety function with the wider tolerance band. The 
remaining revisions to TS 2.1.6 are administrative in nature to change 
defined terms to upper case text. OPPD has also proposed to revise (1) 
item 3 in Table 3-5 of TS 3.2 to require an ``as-left'' PSV lift 
setting tolerance band of +/-1%, and (2) TS 5.9.1c to remove the 
requirement to provide a statement in the Monthly Operating Report 
(MOR) concerning failures or challenges to power operated relief valves 
(PORV) or safety valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The design basis event for RCS over-pressure protection is the 
Loss of Load accident. The Loss of Load event was previously 
evaluated assuming the PSVs lift up to 6% above their setpoint. 
While the proposed amendment widens the tolerance band for installed 
PSVs, only the lower end of the band is changed; therefore, there is 
no adverse affect on the over-pressure protection analysis.
    The proposed amendment does not change the tolerance band 
currently required at the conclusion of PSV surveillance testing 
each refueling outage. As with the current specification, the PSVs 
will continue to be set to within a tolerance band of +/- 1% using 
ASME Code test methods. As a result, the anticipated performance of 
the valves over the course of the subsequent operating cycle is not 
changed. In other words, the potential for setpoint variance exists 
regardless of whether the TSs are changed. The PSVs will begin each 
operating cycle after having been set to open within a lift setting 
tolerance band of +/- 1%. Therefore, the probability or consequences 
of potential setpoint variance during an operating cycle does not 
change. The remaining changes provide supporting statements for the 
wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are 
administrative in nature, or are in accordance with GL 97-02.
    The changes in the case of the defined terms and elimination of 
the TS 5.9.1c Monthly Operating Report concerning failures or 
challenges to PORVs or safety valves are administrative changes 
which do not affect the initiator of an event or prevent safety 
systems from performing their accident mitigation functions as 
assumed in the safety analysis.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Widening the lift setting tolerance band for installed PSVs does 
not create the possibility of a new or different type of accident 
from any previously evaluated.
    The accident analyses address the lift setting tolerance band of 
the PSVs, and the proposed tolerance band does not adversely affect 
the over-pressure protection function and will not compromise RCS 
integrity during power operation. No physical changes to the plant 
are involved.
    The proposed amendment does not change the tolerance band that 
must be met at the conclusion of PSV surveillance testing each 
refueling outage. As with the current Technical Specifications, the 
PSVs will continue to be set at a tolerance band of +/- 1% using 
ASME Code test methods. As a result, the anticipated performance of 
the valves over the course of the subsequent operating cycle is not 
changed. The remaining changes provide supporting statements for the 
wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are 
administrative in nature, or are in accordance with GL 97-02 and 
thus do not create the possibility of a new or different type of 
accident from any previously evaluated.
    The changes in the case of the defined terms and elimination of 
the TS 5.9.1c Monthly Operating Report concerning failures or 
challenges to PORVs or safety valves are administrative changes 
which only affect the technical specifications and do not involve a 
physical change to the plant. Therefore these changes do not alter 
assumptions made in the safety analysis and licensing basis.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Widening the lift setting tolerance band for installed PSVs does 
not involve a significant reduction in a margin of safety. The 
tolerance band of the PSVs is addressed in the accident analyses, 
and the proposed tolerance band does not adversely affect the over-
pressure protection analysis. No physical changes to the plant are 
involved.
    The proposed amendment does not change the tolerance band that 
must be met at the conclusion of PSV surveillance testing each 
refueling outage. As with the current Technical Specifications, the 
PSVs will continue to be set to a tolerance band of +/- 1% using 
ASME Code test methods. As a result, the anticipated performance of 
the valves over the course of the subsequent operating cycle is not 
changed. The remaining changes provide supporting statements for the 
wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are 
administrative in nature, or are in accordance with GL 97-02.
    The changes in the case of the defined terms and elimination of 
the TS 5.9.1c Monthly Operating Report concerning failures or 
challenges to PORVs or safety valves are administrative changes 
which only affect the technical specifications and reporting 
frequency.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: February 19, 2003.
    Description of amendment request: The proposed amendments delete 
requirements from the technical specifications (TS) and other elements 
of the licensing bases to maintain a Post Accident Sampling System 
(PASS). Licensees were generally required to implement PASS upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the TS for nuclear power reactors currently licensed to 
operate. Lessons learned and improvements implemented over the last 20 
years have shown that the information obtained from PASS can be readily 
obtained through other means or is of little use in the assessment and 
mitigation of accident conditions.
    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

[[Page 12958]]

significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on March 20, 
2002 (67 FR 13027). The licensee affirmed the applicability of the 
following NSHC determination in its application dated February 19, 
2003.
    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 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.

    The NRC staff proposes to determine that the amendment requests 
involve 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: Allen G. Howe.

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

    Date of amendment request: December 13, 2003.
    Description of amendment request: The proposed amendment would 
allow the use of Westinghouse leak-limiting Alloy 800 sleeves to repair 
defective steam generator tubes as an alternative to plugging the tube.
    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 in accordance with the three standards set forth in 10 
CFR 50.92(c), which are presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The Westinghouse Alloy 800 leak-limiting repair sleeves are 
designed using the applicable American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code and, therefore, 
meet the design objectives of the original steam generator tubing. 
The applied stresses and fatigue usage for the repair sleeves are 
bounded by the limits established in the ASME Code. Mechanical 
testing has shown that the structural strength of repair sleeves 
under normal, upset, emergency, and faulted conditions provides 
margin to the acceptance limits. These acceptance limits bound the 
most limiting (three times normal operating pressure differential) 
burst margin recommended by NRC's Regulatory Guide 1.121, ``Bases 
for Plugging Degraded PWR Steam Generator Tubes.'' Burst testing of 
sleeve/tube assemblies has demonstrated that no unacceptable levels 
of primary-to-secondary leakage are expected during any plant 
condition.
    The Alloy 800 repair sleeve depth-based structural limit is 
determined using the NRC guidance and the pressure stress equation 
of ASME Code, Section III with additional margin added to account 
for configuration of long axial cracks. A bounding detection 
threshold value has been conservatively identified and statistically 
established to account for growth and determine the repair sleeve/
tube assembly plugging limit. A sleeved tube is plugged on detection 
of degradation in the sleeve/tube assembly.
    Evaluation of the repaired steam generator tube testing and 
analysis indicates no detrimental effects on the sleeve or sleeved 
tube assembly from reactor system flow, primary or secondary coolant 
chemistries, thermal conditions or transients, or pressure 
conditions as may be experienced at Watts Bar Unit 1. Corrosion 
testing and historical performance of sleeve/tube assemblies 
indicates no evidence of sleeve or tube corrosion considered 
detrimental under anticipated service conditions.
    The implementation of the proposed amendment has no significant 
effect on either the configuration of the plant or the manner in 
which it is operated. The consequences of a hypothetical failure of 
the sleeve/tube assembly is bounded by the current steam generator 
tube rupture (SGTR) analysis described in Watts Bar Unit 1 Updated 
Final Safety Analysis Report. Due to the slight reduction in 
diameter caused by the sleeve wall thickness, primary coolant 
release rates

[[Page 12959]]

would be slightly less than assumed for the steam generator tube 
rupture analysis and; therefore, would result in lower total primary 
fluid mass release to the secondary system. A main steam line break 
or feedwater line break will not cause a SGTR since the sleeves are 
analyzed for a maximum accident differential pressure greater that 
that predicted in the Watts Bar Unit 1 safety analysis. The minimal 
repair sleeve/tube assembly leakage that could occur during plant 
operation is well within the Technical Specification leakage limits 
when grouped with current alternate plugging criteria calculated 
leakage values.
    Therefore, TVA has concluded that 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 Alloy 800 leak-limiting repair sleeves are designed 
using the applicable ASME Code as guidance; therefore, it meets the 
objectives of the original steam generator tubing. As a result, the 
functions of the steam generators will not be significantly affected 
by the installation of the proposed sleeve. The proposed repair 
sleeves do not interact with any other plant systems. Any accident 
as a result of potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing SGTR accident 
analysis. The continued integrity of the installed sleeve/tube 
assembly is periodically verified by the Technical Specification 
requirements and the sleeved tube plugged on detection of 
degradation.
    The implementation of the proposed amendment has no significant 
effect on either the configuration of the plant, or the manner in 
which it is operated. Therefore, TVA concludes that this proposed 
change does 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 repair of degraded steam generator tubes with Alloy 800 
leak-limiting repair sleeves restores the structural integrity of 
the degraded tube under normal operating and postulated accident 
conditions and thereby maintains current core cooling margin as 
opposed to plugging the tube and taking it out of service. The 
design safety factors utilized for the repair sleeves are consistent 
with the safety factors in the ASME Boiler and Pressure Vessel Code 
used in the original steam generator design. The portions of the 
installed sleeve/tube assembly that represent the reactor coolant 
pressure boundary can be monitored for the initiation of sleeve/tube 
wall degradation and affected tube plugged on detection. Use of the 
previously identified design criteria and design verification 
testing assures that the margin to safety is not significantly 
different from the original steam generator tubes.
    Therefore, TVA concludes that the proposed change does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: December 13, 2002.
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Plant, Unit 1, Technical Specifications to 
add two new Sections, 3.7.16, ``Shutdown Board Room Air Conditioning 
System,'' and 3.7.17, ``Elevation 772.0 480 Volt Board Room Air 
Conditioning Systems.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration in accordance with the three standards set forth in 10 
CFR 50.92(c), which are presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    [No.] The proposed revision to the [Watts Bar Nuclear Plant] TS 
will provide formalized operational guidance for coping with partial 
or complete unavailability of SDBR [shutdown board room] and 480V 
board room air conditioning (AC) equipment for limited periods of 
time. The change does not impact the frequency of an accident 
because failure of either the SDBR or the 480V board room AC systems 
is not an initiator of any accident scenario. The change does not 
modify any plant hardware including the air conditioning systems, 
and none of their automatic control features or redundant systems 
currently credited in failure analyses are being deleted, modified, 
or otherwise replaced by operator actions as a result of the 
proposed change.
    The proposed TS revision changes current plant operating 
practice and WBN Final Safety Analysis Report (FSAR) assumptions by 
allowing continued power operation with both trains of SDBR air 
conditioning concurrently inoperable and two 480V board room AC 
systems of the same unit to be concurrently inoperable for a limited 
duration, up to 12 hours. This condition is acceptable based on the 
low probability of the occurrence of postulated accidents resulting 
in core damage concurrent with multiple inoperable systems or trains 
of cooling equipment during this timeframe, and based on analyses 
which demonstrate that peak temperatures in each room served by 
these systems remain below mild environment temperature limits 
during this time period. Consequently, there is no significant 
adverse impact on the ability of required safety-related electrical 
equipment to continue to operate and perform their required 
functions, during both normal operation and during design basis 
events. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    [No.] The proposed change does not modify any plant hardware 
including the subject air conditioning systems. The change provides 
specific operational guidance for coping with partial or complete 
unavailability of SDBR and 480V board room air conditioning 
equipment. No new accident or event initiators are created by 
allowing multiple air conditioning systems to be unavailable for the 
limited time period of 12 hours. The supported electrical equipment 
remains capable of performing its intended function both during 
normal operations and post accident. Therefore, the proposed changes 
do 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.
    [No.] The proposed TS revision changes current FSAR assumptions 
by allowing continued power operation with both trains of SDBR air 
conditioning concurrently inoperable and allowing two 480V board 
room air conditioning systems of the same unit to be inoperable for 
a limited duration, up to 12 hours. This condition does not 
significantly reduce the margin of safety due to the low probability 
of the occurrence of a postulated accident resulting in core damage 
concurrent with multiple inoperable systems or trains of cooling 
equipment during the limited time period. In addition, transient 
temperature analyses demonstrate that peak temperatures in each room 
served by these systems remain below mild environment temperature 
limits for a period of 24 hours assuming a complete loss of air 
conditioning to all rooms served by the SDBR and 480V board room AC 
systems concurrently. The analysis is bounding for normal 
operational conditions. Consequently, there is no significant 
adverse impact on the ability of required safety-related electrical 
equipment to continue to operate and perform their required 
functions during both normal operation and during design basis 
events. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

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

[[Page 12960]]

    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 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].

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: September 20, 2002.
    Brief description of amendments: These amendments adopt the generic 
changes approved by Technical Specification Task Force (TSTF) change 
travelers TSTF-349, Revision 1, and TSTF-361, Revision 2, for NUREG-
1430, Revision 1, ``Standard Technical Specifications, Babcock and 
Wilcox Plants,'' dated April 1995, and incorporated into NUREG-1430, 
Revision 2, dated June 2001. Specifically, Section 3.9.5, ``Shutdown 
Cooling (SDC) and Coolant Circulation--Low Water Level,'' is revised to 
add two notes to allow operational changes in the shutdown cooling 
system.
    Date of issuance: February 25, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 256 and 233.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66007).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated February 25, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 5, 2002, as 
supplemented January, 14, 2003.
    Brief description of amendment: The amendment revises the 
surveillance requirements associated with the Containment Isolation 
Valves (CIVs), Reactor Building Closed Cooling Water (RBCCW) System, 
and Service Water (SW) System to remove redundant testing requirements 
that are already addressed by the Inservice Testing Program. Additional 
changes remove the post maintenance testing requirements associated 
with the CIVs, revise the wording of the RBCCW and SW Systems Limiting 
Conditions for Operation, and increase the allowed outage times for the 
RBCCW and SW Systems.
    Date of issuance: February 13, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 273.
    Facility Operating License No. DPR-65: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 16, 2002 (67 FR 
18644). The January 14, 2003, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 14, 2002, as supplemented by letter 
dated December 20, 2002.
    Brief description of amendment: The amendment changes 
administrative Technical Specification 5.5.13 regarding the Containment 
Integrated Leak Rate Testing (ILRT) to allow a one-time extension of 
the interval (to 15 years) for performance of the next ILRT.
    Date of issuance: March 5, 2003.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 131.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42823).
    The December 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 March 5, 2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: December 10, 2002, as 
supplemented on January 20, 2003.
    Brief description of amendment: The Technical Specification (TS) 
amendment request changes the diesel fuel specification to a more 
current revision in TS 4.10.C. The changes also

[[Page 12961]]

make administrative revisions to reflect generic position titles in TS 
6.0; correct page numbers and titles in the Table of Contents; and to 
delete the General Table of Contents. Bases pages were also revised to 
reflect the fuel specification revision, as well as to make 
administrative changes to provide clarity and correct a misspelling.
    Date of Issuance: February 27, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 214.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2802).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated February 27, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 27, 2002.
    Brief description of amendments: The amendments change Appendix B, 
``Environmental Protection Plan,'' of the licensee by removing a 
parenthetical reference to a superseded section of 10 CFR part 51.
    Date of issuance: February 20, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 157/143
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Environmental Protection Plan.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66009).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 20, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of application for amendments: August 7, 2002.
    Brief description of amendments: The amendments: (1) Revised the 
surveillance frequency for air or smoke flow testing of containment 
spray nozzles, as specified in surveillance requirements (SRs) 
4.6.2.1.d and 4.6.2.2.f, from, ``once per 10 years,'' to, ``following 
maintenance which results in the potential for nozzle blockage as 
determined by engineering evaluation;'' (2) allowed the use of a visual 
examination in lieu of an air or smoke flow test; (3) relocated the SR 
4.6.2.2.e.3 criteria for the river/service water flow rate through the 
recirculation spray system heat exchangers to the Updated Final Safety 
Analysis Report; and (4) made minor clarifying changes to the text in 
TS 3.3.1.1.
    Date of issuance: February 24, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 252 and 132.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63694).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 24, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of application for amendments: March 14, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by extending the allowed outage time 
(AOT), or completion time, associated with an inoperable emergency core 
cooling system (ECCS) accumulator. In addition to the AOT extension, 
other changes were incorporated to make the ECCS TSs consistent with 
NUREG-1431, ``Standard Technical Specifications--Westinghouse Plants.'' 
Format and editorial changes were included as necessary to facilitate 
the revision of the TS text to conform to the current TS page format.
    Date of issuance: February 25, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 253 and 133.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21289).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 25, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of application for amendments: October 31, 2002, as 
supplemented by letters dated December 2, 2002, and January 24, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow extending the Type A containment 
integrated leak rate test interval from 10 years to 15 years on a one-
time basis.
    Date of issuance: March 5, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 254 and 134.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75877).
    The December 2, 2002, and January 24, 2003, supplemental letters 
did not change the initial no significant hazards consideration 
determination or expand the amendment beyond the scope of the initial 
notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 5, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: December 9, 2002.
    Brief description of amendment: Pursuant to 10 CFR 50.67, this 
amendment approves the use of Alternative Source Term radiological 
calculations to update the design bases analysis for the Fuel Handling 
Accident as described in the Updated Safety Analysis Report. Regulatory 
Guide 1.183, ``Alternative Radiological Source Terms for Evaluating 
Design-Basis Accidents at Nuclear Power Reactors,'' was used in the 
application.
    Date of issuance: March 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 122.
    Facility Operating License No. NPF-58: This amendment revised the 
Updated Safety Analysis Report.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
804).

[[Page 12962]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 4, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 16, 2002.
    Brief description of amendments: The proposed amendments modified 
Technical Specification (TS) Surveillance Requirement Section 4.0.3 to 
extend the delay time for completion of a missed surveillance to 24 
hours or up to the surveillance frequency, whichever is greater. 
Additionally the proposed change would add a TS Bases Control Program.
    Date of issuance: March 3, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 222 and 217.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 24, 2002 (67 
FR 78521).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 3, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 21, 2002, as 
supplemented by letters dated February 11, 2003, and March 3, 2003.
    Brief description of amendments: The amendments will reduce the 
minimum time required for reactor subcriticality prior to removing 
irradiated fuel from the reactor vessel from 100 hours to 72 hours, as 
specified in Technical Specification 3/4.9.3 ``Refueling Operations, 
Decay Time.''
    Date of issuance: March 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 223 and 218.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68738).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 4, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 11, 2002, as supplemented 
November 11, 2002.
    Brief description of amendments: The amendments would revise the 
Surveillance Requirements for containment leakage rate testing in 
Technical Specification 4.6.1.2 to allow a one-time extension of the 
interval between integrated leakage rate tests from 10 to 15 years.
    Date of issuance: February 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 274 and 254.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34488).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 25, 2003.
    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: February 28, 2001, as supplemented by 
letters dated February 26, September 13 and 27, and November 25, 2002 
(2).
    Brief description of amendment: The amendment consists of changes 
to the design-basis accidents dose assessment methodology and Operating 
License Condition 2.C.(6).
    Date of issuance: February 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 196.
    Facility Operating License No. DPR-46: Amendment revised the final 
safety analysis report and Operating License Condition 2.C.(6).
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48289).
    The supplemental letters provided clarifying information that was 
within the scope of the original Federal Register notice (66 FR 48289) 
and did not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 21, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 3, 2003.
    Brief description of amendment: The amendment changed Technical 
Specifications Surveillance Requirement 3.6.1.7.2 for suppression 
chamber-to-drywell vacuum breaker 2ISC*RV36B to allow an exception to 
the periodic functional testing requirements for the remainder of Cycle 
9.
    Date of issuance: February 21, 2003.
    Effective date: As of the date of issuance to be implemented within 
7 days.
    Amendment No.: 108.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. The Nuclear Regulatory 
Commission published a public notice of the proposed amendment, issued 
a proposed finding of no significant hazards consideration and 
requested that any comments on the proposed no significant hazards 
consideration be provided to the staff by the close of business on 
February 20, 2003. The notice was published in the Syracuse, NY, The 
Post-Standard, on February 11, 2003.
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of New York, and 
final no significant hazards consideration determination are contained 
in a Safety Evaluation dated February 21, 2003.

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

    Date of application for amendment: April 22, 2002, as supplemented 
September 16, 2002.
    Brief description of amendment: The amendment changes the Technical 
Specifications by revising the curves for minimum pressure-temperature 
for the reactor pressure vessel. The P-T curves addressed by this 
amendment were

[[Page 12963]]

developed in accordance with (1) the 1989 edition of the American 
Society of Mechanical Engineers (ASME) Code, section Xl, appendix G, 
(2) 10 CFR part 50, appendix G, and (3) ASME Code Case N-640, 
``Alternative Reference Fracture Toughness for Development of P-T Limit 
Curves.''
    Date of issuance: February 24, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 133.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56323).
    The September 16, 2002, supplemental letter provided additional 
clarifying information that was within the scope of the original 
application, did not change the NRC staff's initial no significant 
hazards consideration determination, and did not expand the scope of 
the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 1, 2002, as supplemented 
November 7, 2002.
    Brief description of amendment: The amendment revises the testing 
frequency for the containment spray nozzles specified in Technical 
Specification Surveillance Requirement 3.6.6.9. The testing frequency 
for the containment spray nozzles is changed from 10 years to 
``following maintenance which could result in nozzle blockage.''
    Date of issuance: February 24, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 211.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63696).
    The November 7, 2002, supplemental letter provided additional 
clarifying information that was within the scope of the original 
application, did not change the NRC staff's initial no significant 
hazards consideration determination, and did not expand the scope of 
the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 8, 2002.
    Brief description of amendment: The amendment relocates the 
requirements of TS 3.5(5) for testing prestressed concrete containment 
tendons to the Fort Calhoun Station, Unit No. 1 Updated Safety Analysis 
Report. The amendment adds the requirement for a Containment Tendon 
Testing Program (TS 5.21) consistent with that presented in Section 5.5 
of NUREG-1432, ``Improved Standard Technical Specification (ITS) for 
Combustion Engineering Plants.''
    Date of issuance: February 26, 2003.
    Effective date: February 26, 2003, and shall be implemented within 
120 days from the date of issuance, including the incorporation of the 
containment tendons testing requirements into the Updated Safety 
Analysis Report.
    Amendment No.: 216.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68741).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 16, 2001, as 
supplemented August 23, 2002, November 8, 2002, and January 20, 2003.
    Brief description of amendments: These amendments revised the 
technical specifications (TSs) to incorporate seven industry-proposed 
Technical Specification Task Force changes (TSTFs) made to NUREG-1433, 
Revision 1, ``Standard Technical Specifications for General Electric 
Plants (BWR/4),'' that have been approved by the Nuclear Regulatory 
Commission.
    Date of issuance: February 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 209 and 183.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64300). The supplements dated August 23, 2002, November 8, 2002, and 
January 20, 2003 provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 25, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 17, 2002, as supplemented 
by letters dated October 30, 2002, December 18, 2002, and January 28, 
2003.
    Brief description of amendments: The amendment revised the values 
of the Safety Limit for Minimum Critical Power Ratio in the Unit 2 
Technical Specifications (TSs) 2.1.1.2, clarified fuel design features 
in TS 4.2.1, and updated the references used to determine the core 
operating limits in TS 5.6.5.b.
    Date of issuance: March 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
upon startup following the Susquehanna Steam Electric Station, Unit 2 
eleventh refueling and inspection outage.
    Amendment Nos.: 184.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53988).
    The supplements dated October 30, 2002, December 18, 2002, and 
January 28, 2003, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 4, 2003.
    No significant hazards consideration comments received: No.

[[Page 12964]]

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: December 2, 2002.
    Brief description of amendments: The amendments revised Technical 
Specification Surveillance Requirement 3.6.4.1.2 to require that only 
one access door in each opening of the secondary containment be closed.
    Date of issuance: February 28, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 236/178.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
812).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2003.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of application for amendments: October 25, 2002, as 
supplemented December 20, 2002, and February 11 and 21, 2003.
    Description of amendment request: The amendment updated the values 
of the Safety Limit Minimum Critical Power Ratio in Technical 
Specification 2.1.1.2 for Cycle 13 operation.
    Date of issuance: February 28, 2003.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 280.
    Facility Operating License No. DPR-52: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75885). The supplemental letters provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original 
request.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 29, 2002, as supplemented 
on October 10, 2002.
    Brief description of amendment: The proposed amendment deletes 
several of the Unit 1 Technical Specification (TS) Surveillance 
Requirements (SR) contained in TS 3/4.4.5, ``Steam Generators'' (SGs), 
associated with the voltage-based SG alternative repair criteria. In 
addition the proposed changes would delete License Condition 2.C.9.d 
which references commitment letters associated with SG inspection 
activities.
    Date of issuance: March 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
during the 2003 Cycle 12 Refueling Outage.
    Amendment No.: 282.
    Facility Operating License No. DPR-77: Amendment revises the TSs.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50960). An October 10, 2002 submittal revised some of the information, 
so a revised notice was published October 29, 2002 (67 FR 66014).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 4, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 10th day of March, 2003.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-6286 Filed 3-17-03; 8:45 am]
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