[Federal Register Volume 69, Number 3 (Tuesday, January 6, 2004)]
[Notices]
[Pages 691-705]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-8]


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UNITED STATES NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 12, 2003, through December 23, 
2003. The last biweekly notice was published on December 23, 2003 (68 
FR 74262).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's 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 February 5, 2004, the licensee may file a request for a hearing 
with respect

[[Page 692]]

to issuance of the amendment to the subject facility operating license 
and any person whose interest may be affected by this proceeding and 
who wishes to participate as a party in the proceeding must file a 
written request for a hearing and a petition for leave to intervene. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested persons 
should consult a current copy of 10 CFR 2.714, which is available at 
the 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.
    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 hearingdocket@nrc.gov. 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 OGCMailCenter@nrc.gov. 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).
    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 
pdr@nrc.gov.

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

    Date of amendment request: December 2, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Surveillance Requirement (SR) 
4.0.2 to extend the delay period, before entering a Limiting Condition 
for Operation, following a missed surveillance. The

[[Page 693]]

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.* * *'' To support this 
change, the following requirement would be added to SR 4.0.2: ``A risk 
evaluation shall be performed for any surveillance delayed greater than 
24 hours and the risk impact shall be managed.'' Additionally, a new 
section 6.2.1 will be added to provide for a TS Bases Control Program.
    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. 
The NSHC determination is restated below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC 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 [a] Margin of Safety

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in [a] 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 [a] 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.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 31, 2003.
    Description of amendment request: The proposed amendment would 
revise the Minimum Critical Power Ratio (MCPR) Safety Limit contained 
in Technical Specification (TS) 2.1.1.2. Currently the MCPR value is 
greater than or equal to 1.12 for two recirculation loop operation and 
greater than or equal to 1.14 for single recirculation loop operation. 
The proposed revised MCPR would be greater than or equal to 1.11 for 
two recirculation loop operation and greater than or equal to 1.12 for 
single recirculation loop operation. Also, a second proposed change 
would add topical report NEDE-32906P-A, ``TRACG Application for 
Anticipated Operational Occurrences (AOO) Transient Analyses,'' to the 
list of documents specified in TS 5.6.5 describing the approved 
methodologies used to determine the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.

Proposed Change 1

    The proposed change to Technical Specification 2.1.1.2 does not 
alter the assumptions of the accident analyses or the Technical 
Specification Bases. The MCPR Safety Limit values are calculated to 
ensure that greater than 99.9 percent of the fuel rods in the core 
avoid transition boiling during any plant operation if the safety 
limit is not violated. The derivation of the MCPR Safety Limit 
values specified in the Technical Specifications has been performed 
using the methods discussed in ``General Electric Standard 
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (i.e., GESTAR-II), 
and U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which 
incorporates Amendment 26. By letters dated November 10, 1999, and 
March 29, 2000, GNF, the NRC approved the use of Amendment 26 to 
NEDE-24011-P-A. Appropriate operational MCPR limits are applied that 
ensure the MCPR Safety Limit is not exceeded during all modes of 
operation and anticipated operational occurrences.
    The revised MCPR Safety Limit values do not affect the 
operability of any plant systems nor do these revised values 
compromise any fuel performance limits; therefore, the probability 
of fuel damage will not be increased as a result of this change. The 
MCPR Safety Limit values do not impact the source term or pathways 
assumed in accidents previously evaluated, and there are no adverse 
effects on the factors contributing to offsite or onsite 
radiological doses. In addition, the revised MCPR Safety Limit 
values do not affect the performance of any equipment used to 
mitigate the consequences

[[Page 694]]

of a previously evaluated accident and do not affect setpoints that 
initiate protective or mitigative actions.
    Therefore, the proposed change to MCPR Safety Limit values 
contained in Technical Specification 2.1.1.2 does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

Proposed Change 2

    The proposed change to TS 5.6.5 will add General Electric 
Nuclear Energy topical report NEDE-32906P-A, ``TRACG Application for 
Anticipated Operational Occurrences (AOO) Transient Analyses,'' to 
the list of documents describing approved methodologies for 
determining core operating limits. Analyzed events are assumed to be 
initiated by the failure of plant structures, systems, or 
components. The core operating limits, which are developed using the 
topical report being added, ensure that the integrity of the fuel 
will be maintained during normal operations and that design 
requirements will continue to be met. The proposed change does not 
involve physical changes to any plant structure, system, or 
component. Therefore, the probability of occurrence for a previously 
analyzed accident is not significantly increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. Use of the analytical methodologies described 
in the topical report being added to TS 5.6.5 will ensure that 
applicable design and safety analyses acceptance criteria are met. 
Use of these NRC-approved methodologies does not affect the 
performance of any equipment used to mitigate the consequences of an 
analyzed accident. As a result, no analysis assumptions are violated 
and there are no adverse effects on the factors that contribute to 
offsite or onsite dose as the result of an accident. Use of the 
approved methodologies described in the topical report being added 
to TS 5.6.5 ensures that plant structures, systems, or components 
are maintained consistent with the safety analysis and licensing 
bases. Based on this evaluation, there is no significant increase in 
the consequences of a previously analyzed event.
    Therefore, the proposed change adding General Electric Nuclear 
Energy topical report NEDE-32906P-A to the TS 5.6.5 list of 
documents describing approved methodologies for determining core 
operating limits 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.

Proposed Change 1

    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation. The proposed revision of the MCPR 
Safety Limit values does not involve installation of any new or 
different equipment. No installed equipment is being operated in a 
different manner than currently evaluated. No new initiating events 
or transients will result from use of the revised MCPR Safety Limit 
values. As a result, no new failure modes are being introduced. 
Therefore, the proposed change to MCPR Safety Limit values contained 
in Technical Specification 2.1.1.2 does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

Proposed Change 2

    The proposed change adding topical report NEDE-32906P-A to TS 
5.6.5, and the use of the analytical methods described therein, does 
not involve any physical alteration of plant systems, structures, or 
components, other than allowing for fuel and core designs in 
accordance with NRC approved methodologies. The proposed methodology 
continues to meet applicable criteria for core operating limit 
analysis. No new or different equipment is being installed. No 
installed equipment is being operated in a different manner. There 
is no alteration to the parameters within which the plant is 
normally operated or in the setpoints that initiate protective or 
mitigative actions. As a result no new failure modes are being 
introduced.
    Therefore, the proposed change adding General Electric Nuclear 
Energy topical report NEDE-32906P-A to the TS 5.6.5 list of 
documents describing approved methodologies for determining core 
operating limits 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.

Proposed Change 1

    The margin of safety is established through the design of the 
plant structures, systems, and components; through the parameters 
within which the plant is operated; through the establishment of 
setpoints for actuation of equipment relied upon to respond to an 
event; and through margins contained within the safety analyses. The 
revised MCPR Safety Limit values will not adversely impact the 
performance of plant structures, systems, components, and setpoints 
relied upon to respond to mitigate an accident or transient. The 
MCPR Safety Limit values are calculated to ensure that greater than 
99.9 percent of the fuel rods in the core avoid transition boiling 
during any anticipated operation occurrences if the safety limit is 
not violated, thereby ensuring that fuel cladding integrity is 
maintained. The revised MCPR Safety Limit values have been 
calculated using NRC approved methods and procedures and preserve 
the existing margin to transition boiling. Based on the assurance 
that the fuel design criteria are being met, the revised MCPR Safety 
Limit values do not involve a reduction in a margin of safety.

Proposed Change 2

    The margin of safety is established through the design of the 
plant structures, systems, and components, through the parameters 
within which the plant is operated, through the establishment of the 
setpoints for the actuation of equipment relied upon to respond to 
an event, and through margins contained within the safety analyses. 
The proposed change adding General Electric Nuclear Energy topical 
report NEDE-32906P-A to the TS 5.6.5 list of documents describing 
approved methodologies for determining core operating limits does 
not impact the condition or performance of structures, systems, 
setpoints, and components relied upon for accident mitigation. The 
proposed change does not significantly impact any safety analysis 
assumptions or results. Therefore, the proposed change adding 
topical report NEDE-32906P-A to the TS 5.6.5 list of documents 
describing approved methodologies for determining core operating 
limits does not result in a significant reduction in the margin of 
safety.
    Based on the above, PEC concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Allen G. Howe.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket 
Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle 
County, Illinois, and Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: November 3, 2003.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) 3.4.1, ``Recirculation Loops 
Operating,'' to add a requirement for the linear heat generation rate 
(LHGR) limits specified in the Core Operating Limits Report (COLR) to 
be met during single recirculation loop operation.
    Technical Specification 3.4.1 for Dresden Nuclear Power Station 
(DNPS) Units 2 and 3, LaSalle County Station

[[Page 695]]

(LSCS) Units 1 and 2, and Quad Cities Nuclear Power Station (QCNPS) 
Units 1 and 2, currently requires limits for average planar linear heat 
generation rate (APLHGR) and minimum critical power ratio (MCPR), as 
well as allowable values for certain Reactor Protection System and 
Control Rod Block functions, to be modified during single recirculation 
loop operation. The modified limits for APLHGR and MCPR are specified 
in the COLR. The proposed change adds a requirement to modify the LHGR 
limit as specified in the COLR with one recirculation loop in 
operation. Although there is currently no TS requirement to adjust the 
LHGR limit during single recirculation loop operation, in accordance 
with NRC Administrative Letter 98-10, ``Dispositioning of Technical 
Specifications that Are Insufficient to Assure Plant Safety,'' 
administrative controls are in place at DNPS and QCNPS to ensure that 
the LHGR limits are appropriately adjusted.
    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 probability or consequences of an accident previously evaluated.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. The LHGR is a measure of the heat generation rate of a 
fuel rod in a fuel assembly at any axial location. Limits on the 
LHGR are specified to ensure that fuel design limits are not 
exceeded anywhere in the core during normal operation, including 
anticipated operational occurrences, and to ensure that the peak 
cladding temperature (PCT) during a postulated design basis LOCA 
does not exceed the limits specified in 10 CFR 50.46.
    LHGR limits have been established consistent with the NRC-
approved GESTAR methodology to ensure that fuel performance during 
normal, transient, and accident conditions is acceptable. The 
proposed change establishes a requirement for LHGR limits to be 
modified, as specified in the COLR, during SLO such that the fuel is 
protected during SLO and during any plant transients or anticipated 
operational occurrences that may occur while in SLO.
    Modifying the LHGR limits during SLO does not increase the 
probability of an evaluated accident. The proposed change does not 
require any physical plant modifications, physically affect any 
plant components, or entail changes in plant operation. Therefore, 
no individual precursors of an accident are affected.
    Limits on the LHGR are specified to ensure that fuel design 
limits are not exceeded anywhere in the core during normal 
operation, including anticipated operational occurrences, and to 
ensure that the PCT during a postulated design basis LOCA does not 
exceed the limits specified in 10 CFR 50.46. This will ensure that 
the fuel design safety criteria (i.e., less than 1% plastic strain 
of the fuel cladding and no fuel centerline melting) are met and 
that the core remains in a coolable geometry following a postulated 
design basis LOCA. Since the operability of plant systems designed 
to mitigate any consequences of accidents has not changed, the 
consequences of an accident previously evaluated are not expected to 
increase.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation. The proposed change does not involve 
any modifications of the plant configuration or allowable modes of 
operation. Requiring the LHGR limits to be modified for SLO by 
applying the SLO LHGR multiplier ensures that the assumptions of the 
LOCA analyses are met.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change will not adversely affect 
operation of plant equipment. The change will not result in a change 
to the setpoints at which protective actions are initiated. LHGR 
limits during SLO are established to ensure that the PCT during a 
postulated design basis LOCA does not exceed the limits specified in 
10 CFR 50.46. This will ensure that the core remains in a coolable 
geometry following a postulated design basis LOCA. The proposed 
change will ensure the appropriate level of fuel protection.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: December 16, 2003.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.4.5, ``Reactor Coolant 
System--Steam Generators,'' to allow a one-time extension of the steam 
generator tube inservice inspection interval from March 2004 to March 
2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their 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 steam generator tubes perform both an accident prevention 
and an accident mitigation function. Steam generator tube integrity 
is necessary to prevent the loss of reactor coolant system inventory 
to the secondary system and to provide a barrier to fission product 
release to the environment. The layup and storage conditions of the 
steam generator during the extended outage have been assessed and 
determined to not adversely affect steam generator conditions. An 
operational assessment of the steam generators for approximately 1.4 
effective full power year has been performed to assure acceptable 
structural integrity during the extended surveillance interval. The 
operational assessment for the steam generators has determined that 
primary-to-secondary leakage following a steam line break, which is 
the limiting event (other than a tube rupture), would continue to be 
acceptable. 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 introduce any new or different 
failure mechanism for the steam generators. Steam generator tube 
integrity will be maintained as previously analyzed following 
postulated events. 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?

[[Page 696]]

    Response: No.
    The layup and storage conditions of the steam generator during 
the extended outage have been assessed and determined to not 
adversely affect steam generator condition. The operational 
assessment for the mid-cycle outage has shown that structural 
margins are greater at approximately 1.4 EFPY then they would be at 
the end of a typical full cycle of operation. Accident induced 
leakage is projected to be the same for the surveillance interval 
extension period as it would be for a full cycle of operation. 
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 E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: January 30, 2003.
    Description of amendment request: This license amendment request 
proposes a revision to the reactor pressure vessel (RPV) material 
surveillance program described within the Perry Nuclear Power Plant 
(PNPP) Updated Safety Analysis Report (USAR) from a plant-specific 
program to the Boiling-Water Reactor Vessel and Internals Project 
(BWRVIP) Integrated Surveillance Program (ISP).
    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.
    NRC [Nuclear Regulatory Commission] regulations impose 
requirements upon the reactor coolant system to ensure that adequate 
safety margins against nonductile or rapidly propagating failures 
exits during normal operation, anticipated operational occurrences, 
and system hydrostatic tests. These requirements are set forth in 10 
CFR 50, Appendix A, ``General Design Criteria for Nuclear Power 
Plants,'' Criterion 31, ``Fracture Prevention of Reactor Coolant 
Pressure Boundary,'' Appendix G, ``Fracture Toughness 
Requirements,'' and Appendix H requires that changes in the fracture 
toughness properties of reactor vessel materials, resulting from the 
neutron irradiation and the thermal environment, are monitored by a 
material surveillance program. To determine the effects of neutron 
fluence on the nil-ductility reference temperature of reactor vessel 
materials, the methods provided in Regulatory Guide (RG) 1.99, 
``Radiation Embrittlement of Reactor Vessel Materials,'' Revision 2 
are used.
    As described in the PNPP USAR, the current PNPP material 
surveillance program is a plant-specific program which complies with 
10 CFR 50, Appendix H.
    The proposed amendment involves changing the material 
surveillance program from a plant-specific program to an integrated 
surveillance program. The use of an integrated program is consistent 
with the requirements of 10 CFR 50, Appendix H, Paragraph III.C. The 
integrated program proposed by PNPP is the BWRVIP ISP. The BWRVIP 
ISP has been reviewed and approved by the NRC staff as an acceptable 
program and is in conformance with 10 CFR 50, Appendix H. Use of the 
ISP, among its many benefits, will increase the number of data 
points used in the evaluation of changes in vessel material 
properties. This will improve compliance with the aforementioned NRC 
regulations. The methods contained in RG 1.99, Revision 2, will 
still be used to determine the effects of neutron fluence upon the 
nil-ductility reference temperature of the PNPP reactor vessel 
materials.
    This change will not affect the reactor pressure vessel, as no 
physical changes are involved. The proposed change will not cause 
the reactor pressure vessel or interfacing systems to be operated 
outside of any design or testing limits. Furthermore, the proposed 
changes will not alter any assumptions previously made in evaluating 
the radiological consequences of any accident. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change revises the PNPP licensing bases to reflect 
participation in the BWRVIP ISP. The ISP was approved by the NRC 
staff as an acceptable material surveillance program which complies 
with 10 CFR 50, Appendix H. The proposed change will not impact the 
manner in which the plant is designed or operated. No new accident 
types or failure modes will be introduced as a result of the 
proposed change. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from that 
previously evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    The material surveillance program requirements contained in 10 
CFR 50, Appendix H, provide assurance that adequate margins of 
safety exist for the reactor coolant system against nonductile or 
rapidly propagating failures during normal operation, anticipated 
operational occurrences, and safety hydrostatic tests. The BWRVIP 
ISP has been approved by the NRC staff as an acceptable material 
surveillance program which complies with 10 CFR 50, Appendix H. The 
ISP will provide the material surveillance data which will ensure 
that the safety margins require by NRC regulations are maintained 
for the PNPP reactor coolant system. Therefore, the proposed change 
does not involve a significant reduction in any margins 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 E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: August 14, 2003.
    Description of amendment request: This license amendment request 
(LAR) proposes a revision to increase the analytical limit and the 
resulting Technical Specification (TS) allowable value (AV) related to 
the setpoint for the Main Steam Line Turbine Building Temperature--
High, system isolation function. This LAR revises the main steam line 
trip setpoint AV based on improved computer modeling of the expected 
building temperature transients in the event of a larger steam leak. 
The proposed change improves the operating margins and reduces 
challenges to the plant by avoiding unnecessary plant shutdown 
transients from turbine building high temperatures from other than a 
main steam line leak.
    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 nuclear boiler Leak Detection System (LDS) instrumentation 
associated with the proposed amendment assists in the detection of a 
small steam leak to prevent a significant release of radioactive 
material created by conditions other than a break within the Reactor 
Coolant Pressure Boundary (RCPB).
    The proposed amendment establishes a new steam leak system 
isolation temperature limit in the Turbine Building.

[[Page 697]]

    There is no accident analysis or transient that credits the 
subject LDS instrumentation. The subject instrumentation is for the 
detection of small steam leaks and not a pipeline break as described 
in the Updated Safety Analysis Report (USAR) Chapter 15 accident 
analysis. The detection of main steam line flow is the parameter 
used in the accident analysis to signal a steam line break outside 
of containment.
    The proposed amendment does not impact the physical design or 
location of the LDS instrumentation. This proposed amendment is 
associated only with the results of a main steam line leak in the 
non-safety related Turbine Building and has no impact on the 
initiation of this leak. The analysis completed in support of the 
proposed amendment indicates that the radiological effects 
associated with the new steam leak system isolation limit remains 
bounded by the existing large main steam line break analysis 
contained within the PNPP [Perry Nuclear Power Plant] USAR. The 
proposed leakage limit does not alter the current function of the 
LDS that isolates the Main Steam system prior to the leakage 
degrading to a point where the system integrity, i.e., piping 
integrity and makeup capability, is challenged. Therefore, the 
proposed amendment ensures that the criteria for acceptance as 
established in the original licensing bases and the requirements of 
the original design basis remain valid. It has been determined that 
the service life, i.e., Equipment Qualification (EQ) and structural 
integrity of the Structures, Systems and Components (SSC) in the 
affected areas are not adversely impacted by the proposed amendment.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed amendment does not impact the physical design or 
location of the associated LDS instrumentation. The instruments will 
still promptly initiate the automatic isolation of the appropriate 
Containment and Drywell isolation valves to mitigate steam leakage 
as credited in the original licensing bases. This proposed amendment 
is associated only with the results of a main steam line break in 
the non-safety related Turbine Building and has no impact on the 
initiation of this leak. The analysis completed in support of the 
proposed amendment indicates that the radiological effects 
associated with the new steam leak system isolation limit remains 
bounded by the existing large main steam line break analysis 
contained within the PNPP USAR. The EQ and structural integrity of 
any SSC located within the non-safety related Turbine Building are 
not affected by the proposed amendment. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed change will not involve a single reduction in 
the margin of safety.
    The analysis performed for the proposed amendment proves that 
the appropriate instruments will still promptly initiate automatic 
system isolation, upon sensing temperatures in excess of their 
setpoints. The radiological effects associated with the proposed 
small steam leak to be detected remain bounded by the existing large 
main steam line break analysis contained within the USAR. Steam 
leaks in the affected area of the Turbine Building will be detected 
on a timely basis so that the Main Steam system will be isolated 
before such degradation could become sufficiently severe to 
jeopardize the safety of the system. Also, steam leaks will be 
detected before the leakage could increase to a level beyond the 
capability of the makeup system. Therefore, the proposed amendment 
ensures that the criteria for acceptance as established in the 
original licensing bases and the requirements of the original design 
basis remain valid. There is no accident analysis or transient that 
credits the associated leak detection instrumentation, and the LDS 
Main Steam Line Turbine Building Temperature--High function is 
categorized as non-risk significant. Further, the proposed amendment 
reduces the challenges to SSCs due to unnecessary plant shutdowns 
created by conditions other than a main steam line leak. The EQ and 
structural integrity of any SSC located within the Turbine Building 
are not affected by the proposed amendment. Therefore, the proposed 
amendment 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 29, 2003.
    Description of amendment request: The proposed license amendments 
would allow relocation of specific pressure and flow values for the 
boric acid makeup (BAM) pumps, containment spray (CS) pumps, high 
pressure safety injection (HPSI) pumps, and low pressure safety 
injection (LPSI) pumps from the St. Lucie Units 1 and 2 Technical 
Specifications to the Updated Final Safety Analysis Reports (UFSARs). 
This is consistent with the Combustion Engineering Improved Standard 
Technical Specifications and the Nuclear Regulatory Commission Final 
Policy Statement on Technical Specification Improvements for Nuclear 
Power Reactors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Would operation of the facility in accordance with the 
proposed amendments involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to relocate the BAM, CS, HPSI, and LPSI 
pump surveillance verification details in the aforementioned 
Technical Specifications surveillance requirements to the St. Lucie 
UFSARs do not adversely affect accident initiators or precursors nor 
alter the design assumptions, conditions, configuration of the 
facility, or the manner in which it is operated. The proposed 
changes do not alter or prevent the ability of structures, systems, 
or components to perform their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the St. Lucie UFSARs.
    The subject surveillance requirement criteria relocated to the 
St. Lucie UFSARs will continue to be administratively controlled. 
Changes to the St. Lucie UFSARs are evaluated and controlled under 
10 CFR 50.59 prior to implementation. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) Would operation of the facility in accordance with the 
proposed amendments create the possibility of a new different kind 
of accident from 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 plant is operated.
    There are no changes to the source term or radiological release 
assumptions used in evaluating the radiological consequences in the 
St. Lucie UFSARs. The proposed changes have no adverse impact on the 
component or system interactions. The proposed changes will not 
adversely degrade the ability of systems, structures and components 
important to safety to perform their safety function nor change the 
response of any system, structure or component important to safety 
as described in the UFSARs. The proposed changes do not change the 
level of programmatic and procedural details of assuring operation 
of the facility in a safe manner. Since there are no changes to the 
design assumptions, conditions, configuration of the facility, or 
the manner in which the plant is operated and surveilled, the 
proposed changes do not create the possibility of a new different 
kind of accident from any previously analyzed.
    (3) Would operation of the facility in accordance with the 
proposed amendments involve a significant reduction in a margin of 
safety?

[[Page 698]]

    There is no adverse impact on equipment design or operation and 
there are no changes being made to the Technical Specification 
required safety limits or safety system settings that would 
adversely affect plant safety. The proposed changes do not reduce 
the level of programmatic or procedural controls associated with the 
activities presently performed via the aforementioned surveillance 
requirements.
    Future changes to the relocated technical requirements will 
require an evaluation pursuant to the provisions of 10 CFR 50.59 
prior to implementation.
    Therefore, relocation of the specific pump pressure and flow 
criteria contained in the aforementioned Technical Specification 
Surveillance Requirements to the St. Lucie Units 1 and 2 UFSARs does 
not involve a significant reduction in the margin of safety provided 
in the existing specifications.

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

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

    Date of amendment request: November 21, 2003.
    Description of amendment request: The proposed amendments would 
transfer Technical Specification (TS) requirements 6.5 (Review and 
Audit), 6.8.2 and 6.8.3 (procedures and programs review specifics), and 
6.10 (Record Retention) to the quality assurance plan (a licensee 
controlled document) for St. Lucie Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to the St. Lucie Plant TS do not adversely 
affect accident initiators or precursors, nor alter the design 
assumptions, conditions, and configuration of the facility or the 
manner in which the plant is operated and maintained. In addition, 
the proposed changes do not affect the manner in which the plant 
responds in normal operation, transient, or accident conditions, nor 
do they change any of the procedures related to operation of the 
plant. The proposed changes do not alter or prevent the ability of 
structures, systems, and components (SSCs) to perform their intended 
function to mitigate the consequences of an initiating event within 
the acceptance limits assumed in the Updated Final Safety Analysis 
Report (UFSAR). The proposed changes are administrative for the 
purpose of updating TS to reflect current NRC and industry 
initiatives.
    The proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated in 
the St. Lucie UFSARs. Further, the proposed changes do not increase 
the types and amounts of radioactive effluent that may be released 
off site, nor significantly increase individual or cumulative 
occupational/public radiation exposures.
    Therefore, it is concluded that these proposed revisions do not 
involve a significant increase in the probability or consequence of 
an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes to the St. Lucie Plant TS do not change the 
operation or the design basis of any plant system or component 
during normal or accident conditions. The proposed changes do not 
include any physical changes to the plant. In addition, the proposed 
changes do not change the function or operation of plant equipment 
or introduce any new failure mechanisms. The plant equipment will 
continue to respond per the design and analyses and there will not 
be a malfunction of a new or different type introduced by the 
proposed changes.
    The proposed changes are administrative in nature and only 
correct, update and clarify the St. Lucie Plant Technical 
Specifications to reflect NRC guidance, i.e., AL 95-06. The proposed 
changes do not modify the facility nor do they affect the plant's 
response to normal, transient, or accident conditions. The changes 
do not introduce a new mode of plant operation. The changes are an 
enhancement and do not affect plant safety. The plant's design and 
design basis are not revised and the current safety analyses remains 
in effect.
    Thus, these proposed revisions to the St Lucie Plant TS do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes are administrative changes to the St. Lucie 
Plant Technical Specifications. The safety margins established 
through Limiting Conditions for Operation, Limiting Safety System 
Settings and Safety Limits as specified in the Technical 
Specifications are not revised nor is the plant design or its method 
of operation revised by the proposed changes.
    Thus, it is concluded that these proposed revisions to the St. 
Lucie Plant TS do not involve a significant reduction in a margin of 
safety.

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

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

    Date of amendment request: December 6, 2003.
    Description of amendment request: The proposed amendment would 
revise the Unit 1 and 2 Technical Specifications (TSs) by adding a 
requirement to apply linear heat generation rate [LHGR] limits if the 
main turbine bypass system becomes inoperable.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change to the TS 3.7.6 does not directly or 
indirectly affect any plant system, equipment, component, or change 
the processes used to operate the plant. Further, the MCPR [minimum 
critical power ratio] and LHGR limits documented in the unit/cycle 
specific COLRs [core operating limits report] for Main Turbine 
Bypass System operable and inoperable are generated using NRC 
[Nuclear Regulatory Commission] approved methodology and meet the 
applicable acceptance criteria. The COLR operating limits thus 
assure that the MCPR Safety Limit and LHGR Limit will not be 
exceeded during normal operation or anticipated operational 
occurrences. Thus, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to TS 3.7.6 does not directly or indirectly 
affect any plant system, equipment, or component and therefore does 
not affect the failure modes of any of these items. Thus, the 
proposed changes do not create the possibility of a previously 
unevaluated operator error or a new single failure.

[[Page 699]]

    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Since the proposed changes do not alter any plant system, 
equipment, component, or the processes used to operate the plant, 
the proposed change will not jeopardize or degrade the function or 
operation of any plant system or component governed by Technical 
Specifications. The proposed change to TS 3.7.6 does not involve a 
significant reduction in the margin of safety as currently defined 
in the Bases of the applicable Technical Specification sections, 
because the MCPR and LHGR limits calculated for Main Turbine Bypass 
System operable and inoperable preserve the required margin of 
safety.
    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: Bryan A. Snapp, Esquire, Assoc, General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101,1179.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 22, 2003 (TSC 03-12).
    Description of amendment request: The proposed change involves the 
extension from 1 hour to 24 hours of the completion time (CT) for 
Action (a) of Technical Specification (TS) 3.5.1.1, which defines 
requirements for accumulators. Accumulators are part of the emergency 
core cooling system and consist of tanks partially filled with borated 
water and pressurized with nitrogen gas. The contents of the tank are 
discharged to the reactor coolant system (RCS) if, as during a loss-of-
coolant accident, the coolant pressure decreases to below the 
accumulator pressure. Action (a) of TS 3.5.1.1 specifies a CT to 
restore an accumulator to operable status when it has been declared 
inoperable for a reason other than the boron concentration of the water 
in the accumulator not being within the required range. This change was 
proposed by the Westinghouse Owners Group participants in the TS Task 
Force (TSTF) and is designated TSTF-370. TSTF-370 is supported by NRC-
approved topical report WCAP-15049-A, ``Risk-Informed Evaluation of an 
Extension to Accumulator Completion Times,'' submitted on May 18, 1999. 
The NRC staff issued a notice of opportunity for comment in the Federal 
Register on July 15, 2002 (67 FR 46542), on possible amendments 
concerning TSTF-370, 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 March 12, 
2003 (68 FR 11880). The licensee affirmed the applicability of the 
following NSHC determination in its application dated October 22, 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 basis for the accumulator limiting condition for operation 
(LCO), as discussed in Bases Section 3.5.1.1, is to ensure that a 
sufficient volume of borated water will be immediately forced into 
the core through each of the cold legs in the event the RCS pressure 
falls below the pressure of the accumulators, thereby providing the 
initial cooling mechanism during large RCS pipe ruptures. As 
described in Section 9.2 of WCAP-15049-A, the proposed change will 
allow plant operation with an inoperable accumulator for up to 24 
hours, instead of 1 hour, before the plant would be required to 
begin shutting down. The impact of the increase in the accumulator 
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total 
plant core damage frequency (CDF) less than 1.0E-03/yr. The 
incremental conditional core damage probabilities calculated in 
WCAP-15049-A for the accumulator CT increase meet the criterion of 
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using 
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' for all cases except those that are based on 
design basis success criteria. As indicated in WCAP-15049-A, design 
basis accumulator success criteria are not considered necessary to 
mitigate large break loss-of-coolant accident (LOCA) events, and 
were only included in the WCAP-15049-A evaluation as a worst case 
data point. In addition, WCAP-15049-A states that the NRC has 
indicated that an incremental conditional core damage frequency 
(ICCDP) greater than 5E-07 does not necessarily mean the change is 
unacceptable.
    The proposed technical specification change does not involve any 
hardware changes nor does it affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, engineered safety feature (ESF) actuation setpoints, 
accident mitigation capabilities, accident analysis assumptions or 
inputs.
    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

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this 
proposed technical specification CT increase. All safety systems 
still function in the same manner and there is no additional 
reliance on additional systems or procedures. The proposed 
accumulator CT increase has a very small impact on core damage 
frequency. The WCAP-15049-A evaluation demonstrates that the small 
increase in risk due to increasing the CT for an inoperable 
accumulator is within the acceptance criteria provided in RGs 1.174 
and 1.177. No new accidents or transients can be introduced with the 
requested change and the likelihood of an accident or transient is 
not impacted.
    The malfunction of safety related equipment, assumed to be 
operable in the accident analyses, would not be caused as a result 
of the proposed technical specification change. No new failure mode 
has been created and no new equipment performance burdens are 
imposed.
    Therefore, 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 proposed change does not involve a significant reduction in 
a margin of safety. There will be no change to the departure from 
nucleate boiling ratio (DNBR) correlation limit, the design DNBR 
limits, or the safety analysis DNBR limits.
    The basis for the accumulator LCO, as discussed in Bases Section 
3.5.1.1, is to ensure that a sufficient volume of borated water will 
be immediately forced into the core through each of the cold legs in 
the event the RCS pressure falls below the pressure of the 
accumulators, thereby providing the initial cooling mechanism during 
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable 
accumulator for up to 24 hours, instead of 1 hour, before the plant 
would be required to begin shutting down. The impact of this on 
plant risk was evaluated and found to be very small. That

[[Page 700]]

is, increasing the time the accumulators will be unavailable to 
respond to a large LOCA event, assuming accumulators are needed to 
mitigate the design basis event, has a very small impact on plant 
risk.
    Since the frequency of a design basis large LOCA (a large LOCA 
with loss of offsite power) would be significantly lower than the 
large LOCA frequency of the WCAP-15049-A evaluation, the impact of 
increasing the accumulator CT from 1 hour to 24 hours on plant risk 
due to a design basis large LOCA would be significantly less than 
the plant risk increase presented in the WCAP-15049-A evaluation.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of application request: December 8, 2003.
    Description of amendment request: The licensee is proposing to 
revise Technical Specification (TS) Section 5.5.6, ``Containment Tendon 
Surveillance Program,'' for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The proposed 
revision to TS 5.5.6 is to indicate that the Containment Tendon 
Surveillance Program, inspection frequencies, and acceptance criteria 
shall be in accordance with Section XI, Subsection IWL of the ASME 
Boiler and Pressure Vessel Code and the applicable addenda as required 
by 10 CFR 50.55a, except where an exemption or relief has been 
authorized by the NRC. The licensee has also proposed to delete the 
provisions of Surveillance Requirement 3.0.2 from this TS. In addition, 
the licensee is proposing to revise TS 5.5.16, ``Containment Leakage 
Rate Testing Program,'' to add exceptions to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Testing Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The revised 
requirements do not affect the function of the containment post-
tensioning system components. The post-tensioning systems are 
passive components whose failure modes could not act as accident 
initiators or precursors.
    The proposed change affects the frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Containment Leakage Rate Testing Program. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The frequency of visual examinations of the concrete 
surfaces of the containment and the mode of operation during which 
those examinations are performed has no relationship to or adverse 
impact on the probability of any of the initiating events assumed in 
the accident analyses. The proposed change would allow visual 
examinations that are performed pursuant to NRC approved ASME 
Section XI Code requirements (except where relief has been granted 
by the NRC) to meet the intent of visual examinations [as] required 
by Regulatory Guide 1.163, [because of the commitment in Appendix 3A 
of the Callaway Final Safety Analysis Report,] without requiring 
additional visual examinations pursuant to the Regulatory Guide. The 
intent of early detection of deterioration will continue to be met 
by the more rigorous requirements of the Code required visual 
examinations. As such, the safety function of the containment as a 
fission product barrier is maintained.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The 
function of the containment post-tensioning system components are 
not altered by this change. The change affects the frequency of 
visual examinations that will be performed for the concrete surfaces 
[of the containment]. In addition, the proposed change allows those 
examinations to be performed during power operation as opposed to 
during a refueling outage. The proposed change does not involve a 
modification to the physical configuration of the plant (i.e., no 
new equipment will be installed) or change in the methods governing 
normal plant operation. The proposed change will not impose any new 
or different requirements or introduce a new accident initiator, 
accident precursor, or malfunction mechanism. Additionally, there is 
no change in the types or increases in the amounts of any 
effluent[s] that may be released off-site and there is no increase 
in individual or cumulative occupational exposure.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The 
function of the containment post-tensioning system components are 
not altered by this change. The change affects the frequency of 
visual examinations that will be performed for the concrete surfaces 
[of the containment]. In addition, the proposed change allows those 
examinations to be performed during power operation as opposed to 
during a refueling outage. The safety function of the containment as 
a fission product barrier will be 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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 30, 2003.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) for alternating current (AC) sources--
operating (TS 3.8.1) and electrical power distribution systems--
operating (TS 3.8.9) by extending the required action completion times 
(CTs). For TS 3.8.1, the amendment would extend the CT to restore a 
single inoperable diesel generator (DG) to operable status by adding a 
note to the CT for Required Action B.4. A note would also be added to 
the CT for Required Action A.3 to restore a single inoperable offsite 
circuit to operable status to account for the note that would be added 
to the CT for Required Action B.4.
    For TS 3.8.9, the CT for Required Action C.1 (to restore a single 
inoperable AC vital bus subsystem to

[[Page 701]]

operable status) would be extended to 24 hours. The second CTs, from 
the discovery of the failure to meet the limiting condition for 
operation (LCO), for Required Actions B.1 (to restore a single 
inoperable AC electrical power distribution subsystem to operable 
status), C.1 (given above), and D.1 (to restore a single inoperable 
direct current (DC) electrical power distribution subsystem to operable 
status) would be extended to 34 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to the Completion Times do not change the 
response of the plant to any accidents and have an insignificant 
impact on the reliability of the electrical power sources and 
distribution systems. The proposed changes to the second Completion 
Times are administrative in nature and only intended to prevent the 
plant from successively entering and exiting ACTIONS associated with 
different systems governed by one LCO without ever meeting the LCO. 
The electrical power sources and distribution subsystems will remain 
highly reliable and the proposed changes 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 core damage frequency (CRF) and large early release frequency 
(LERF) is acceptable. In addition, for the Completion Time change, 
the incremental conditional core damage probabilities (ICCDP) and 
incremental conditional large early release probabilities (ICLERP) 
are also acceptable. These changes are consistent with the 
acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, 
since the electrical sources and distribution subsystems will 
continue to perform their [safety] functions with high reliability 
as originally assumed, and the increase in risk as measured by CDF, 
LERF, ICCDP, [and] ICLERP is acceptable, there will not be a 
significant increase in the consequences of any accidents 
[previously evaluated].
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended [safety] function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed changes do not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Further, the proposed changes do 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 changes are consistent with the 
safety analysis assumptions and resultant [radiological] 
consequences.
    Therefore, the proposed change[s do] not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not result in a change in the manner in 
which the electrical distribution subsystems provide plant 
protection. The use of the Sharpe Station will provide an alternate 
AC power source in the event of emergent inoperability of a WCGS 
[Wolf Creek Generating Station] DG or a complete loss of all WCGS 
emergency AC power. The changes do not alter assumptions made in the 
safety analysis. The changes to Completion Times do not change any 
existing accident scenarios, nor create any new or different 
accident scenarios. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed change[s do] not create the possibility 
of a new or different [kind of] accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not alter the 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 these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis. The calculated impact on risk is insignificant and is 
consistent with the acceptance criteria contained in Regulatory 
Guides 1.174 and 1.177.
    Therefore, the proposed change[s do] not involve a significant 
reduction in a margin of safety.

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

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 pdr@nrc.gov.

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

    Date of application for amendment: November 8, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications to delete the requirements for the auxiliary and fuel 
handling building air treatment system.
    Date of issuance: December 12, 2003.

[[Page 702]]

    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 248.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 24, 2002 (67 
FR 78517).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 12, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 28, 2003, as supplemented 
October 8, 2003.
    Brief description of amendment: This amendment eliminates the need 
to credit Boraflex neutron-absorbing material for reactivity control in 
the spent fuel storage pool.
    Date of issuance: December 22, 2003.
    Effective date: December 22, 2003.
    Amendment No.: 198.
    Facility Operating License No. DPR-23: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40710). The October 8, 2003, supplement contained clarifying 
information only and did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 11, 2002, as 
supplemented June 24, 2003.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) related to N-1 loop operation. Specifically, the 
changes eliminate N-1 loop operation from particular sections of the 
TSs and makes other changes that are clarifying and/or administrative 
in nature. In addition, the TS Bases are revised to address the 
proposed changes.
    Date of issuance: December 10, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 217.
    Facility Operating License No. NPF-49: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2800). The January 24, 2003, supplement contained clarifying 
information and did not change the staff's proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 10, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 24, 2003, as supplemented 
by letters dated June 25 and October 15, 2003.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to relocate certain reactor coolant 
system cycle-specific parameter limits from the TSs to the Core 
Operating Limits Report, and revises the minimum allowable reactor 
coolant system flow rate.
    Date of issuance: December 19, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 210 and 204.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54749), November 18, 2003 (68 FR 65090).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: May 19, 2003, as supplemented 
October 27, 2003.
    Brief description of amendment: This amendment revises the 
Technical Specifications to require ``flow indication,'' rather than 
``safety-grade flow indication,'' to satisfy Surveillance Requirement 
4.7.1.7.e.2 for the motor driven feedwater pump.
    Date of issuance: December 18, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 261.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34669).
    The supplement dated October 27, 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 amendment is contained in a 
Safety Evaluation dated December 18, 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: July 15, 2003.
    Brief description of amendment: The amendment revises the Technical 
Specification (TS) requirements for surveillance of the status of 
Secondary Containment Isolation Valves and Blind Flanges in 
Surveillance Requirement 3.6.4.2.1, consistent with TS Task Force 
Traveler-45 Revision 2.
    Date of issuance: December 5, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 202.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 28, 2003 (68 FR 
61479).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 5, 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: April 11, 2003.
    Brief description of amendment: The amendment revises TS Section 
5.0, ``Administrative Controls,'' to make various administrative, 
editorial, and typographical changes.
    Date of issuance: December 15, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.

[[Page 703]]

    Amendment No.: 213.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64136).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 15, 2003.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit 1, San Diego County, California

    Date of application for amendment: July 25, 2003, supplemented by 
letters dated October 3, 2003, and December 3, 2003.
    Brief description of amendment: This amendment approves the use of 
the modified Unit 1 turbine gantry crane and turbine building support 
structure in a single failure proof application and at a rated capacity 
of 105 tons for handling of spent fuel casks as documented in the 
Defueled Safety Analysis Report (DSAR). The DSAR changes approved by 
this amendment are needed to permit use of the modified turbine gantry 
crane and turbine building support structure for lifting and handling 
of the spent fuel casks from the SONGS Unit 1 spent fuel pool to the 
Independent Spent Fuel Storage Installation.
    Date of issuance: December 18, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: Unit 1-162.
    Facility Operating License No.DPR-13: Amendment revises the license 
to permit use of the turbine building gantry crane in a single failure 
proof application at a rated capacity of 105 tons for handling of spent 
fuel casks.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54751).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 18, 2003.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: December 19, 2002, as 
supplemented October 20, 2003.
    Brief description of amendments: These amendments correct various 
typographical, editorial, and other administrative errors currently in 
the Technical Specifications for Surry Power Station, Units 1 and 2.
    Date of issuance: December 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 238 and 237.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5684). The supplement dated October 20, 2003, provided clarifying 
information only and did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 16, 2003.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety

[[Page 704]]

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 Assess 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 Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to pdr@nrc.gov.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By February 5, 2004, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and 
electronically on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If there are problems in accessing the 
document, contact the PDR Reference staff at 1-800-397-4209, 301-415-
4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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: Rulemakings 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 the 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 hearingdocket@nrc.gov. A copy of the petition for 
leave to intervene and request for hearing 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 OGCMailCenter@nrc.gov. 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagorda County, Texas

    Date of amendment request: December 23, 2003.
    Description of amendment request: The amendments revise Technical 
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to extend the 
allowed outage time for Unit 2 Standby Diesel Generator 22 from 14 days 
to 21 days as a one-time change for the purpose of collecting data 
associated with failure of SDG-22.
    Date of issuance: December 23, 2003.
    Effective date: December 23, 2003.
    Amendment Nos.: Unit No. 2: 148.
    Facility Operating License Nos. NPF-76 and NPF-80: Amendments 
revise the Technical Specifications.

[[Page 705]]

    Public comments requested as to final no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated December 23, 
2003.
    Attorney for licensee: A. H. Gutterman, Esquire, Morgan, Lewis & 
Bockius, LLP, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

    Dated at Rockville, Maryland, this 24th day of December, 2003.
    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-8 Filed 1-5-04; 8:45 am]
BILLING CODE 7590-01-P