[Federal Register Volume 69, Number 41 (Tuesday, March 2, 2004)]
[Notices]
[Pages 9857-9871]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-4343]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, February 5, 2004, through February 19, 
2004. The last biweekly notice was published on February 17, 2004 (69 
FR 7517).

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

[[Page 9858]]

determination. Within 60 days after the date of publication of this 
notice, 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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place 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 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, 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 2.309, 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 a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor 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.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV or 
(4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966.

[[Page 9859]]

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 it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by email 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 requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    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 email to 
pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois, Docket No. 50-219, Oyster Creek 
Generating Station, Ocean County, New Jersey, Three Mile Island Nuclear 
Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: January 30, 2004.
    Description of amendment request: The licensee proposes to revise 
the operating licenses to reflect the current 100% ownership of AmerGen 
by Exelon Generation Company. In particular, the proposed amendments 
will remove PECO and British Energy from the licenses, and will remove 
certain license conditions in their entirety which were imposed to 
acknowledge the indirect foreign ownership in AmerGen by British Energy 
plc. Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes are administrative in nature and would 
merely conform the facility operating licenses to reflect the 
current ownership structure of AmerGen. No actual plant equipment or 
accident analyses will be affected by the proposed changes. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature and would 
merely conform the facility operating licenses to reflect the 
current ownership structure of AmerGen. No actual plant equipment or 
accident analyses will be affected by the proposed changes and no 
failure modes not bounded by previously evaluated accidents will be 
created.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is administrative in nature and would merely 
conform the facility operating licenses to reflect the current 
ownership structure of AmerGen. No actual plant equipment or 
accident analyses will be affected by the proposed changes. 
Additionally, the proposed changes will not relax any criteria used 
to establish safety limits, will not relax any safety system 
settings, or will not relax the bases for any limiting conditions 
for operation. Therefore, the proposed changes do not involve a 
significant reduction in any margin of safety.

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

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2 (IP2), Westchester County, New York

    Date of amendment request: January 29, 2004.
    Description of amendment request: The proposed amendment would 
increase the maximum authorized reactor core power level from 3114.4 
megawatt thermal (MWt) to 3216 MWt. This represents a nominal increase 
of 3.26% rated thermal power. Basis for proposed no significant hazards 
consideration determination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The evaluations and analyses associated with this proposed 
change to core power level have demonstrated that all applicable 
acceptance criteria for plant systems, components, and analyses 
(including the Final Safety Analysis Report Chapter 14 safety 
analyses) will continue to be met for the proposed increase in 
licensed core thermal power for IP2. The subject increase in core 
thermal power will not result in conditions that could adversely 
affect the integrity (material, design, and construction standards) 
or the operational performance of any potentially affected system, 
component or analysis. Therefore, the probability of an accident 
previously evaluated is not affected by this change. The subject 
increase in core thermal power will not adversely affect the ability 
of any safety-related system to meet its intended safety function. 
Further, the radiological dose evaluations in support of this power 
uprate effort show all acceptance criteria are met.
    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 evaluations of this proposed amendment show that all 
applicable acceptance criteria for plant systems, components, and 
analyses (including FSAR [final safety analysis report] Chapter 14 
safety analyses) will continue to be met for the proposed power 
increase in IP2 licensed core thermal power. The subject increase in 
core thermal power will not result in conditions that could 
adversely affect the integrity (material, design, and construction 
standards) or operational performance of any potentially affected 
system, component, or analyses. The subject increase in core thermal 
power will not adversely affect the ability of any safety-related 
system to meet its safety function. Furthermore, the conditions and 
changes associated with the subject increase in core thermal power 
will neither cause initiation of any accident, nor create any new 
credible limiting single failure. The power uprate does not result 
in changing the status of events previously deemed to be non-
credible being made credible. Additionally, no new operating modes 
are proposed for the plant as a result of this requested change.
    Therefore, the subject increase in core thermal power level will 
not create the

[[Page 9860]]

possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The evaluations associated with this proposed change show that 
all applicable acceptance criteria for plant systems, components, 
and analyses (including FSAR Chapter 14 safety analyses) will 
continue to be met for this proposed increase in IP2 licensed core 
thermal power. The subject increase in core thermal power will not 
result in conditions that could adversely affect the integrity 
(material, design, and construction standards) or operational 
performance of any potentially affected system, component, or 
analysis. The subject power uprate will not adversely affect the 
ability of any safety-related system to meet its intended safety 
function.
    Therefore, the subject increase in core thermal power will not 
involve a significant reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: February 9, 2004.
    Description of amendment request: The proposed amendment would 
remove the pressurizer heatup and cooldown limits, and the associated 
action and surveillance requirements, from the Technical Specifications 
and place them in a licensee controlled document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The probability of an accident is unchanged as a result of the 
proposed change to delete the ANO-2 [Arkansas Nuclear One, Unit 2] 
pressurizer heatup and cooldown rates and associated action, 
surveillance requirement, and bases from the TS [Technical 
Specification]. The cooldown and heatup rates are not initiators to 
any accidents or pressurizer transients discussed in the ANO-2 SAR 
[Safety Analysis Report]. Therefore, the probability of an accident 
is not changed.
    The purpose of the pressurizer heatup and cooldown limits is to 
ensure that given transient events will not negatively affect the 
pressurizer structural integrity beyond Code [American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code] 
allowables. These limits will be maintained within ASME Code 
allowables in a licensee controlled document in accordance with 10 
CFR 50.59. Therefore, the consequences of an accident are not 
increased.
    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 limitations imposed on the pressurizer heatup and cooldown 
rates are provided to assure that the pressurizer is operated within 
the design criteria assumed for the flaw evaluation and fatigue 
analysis performed in accordance with the ASME Code Section XI, 
subsection IWB-3600 requirements. The ANO-2 SAR has analyzed the 
conditions that would result from a thermal or pressurization 
transient on the ANO-2 pressurizer. The proposed deletion of the 
pressurizer heatup and cooldown rates and relocation of the limits 
to a licensee controlled document does not change the way that the 
pressurizer is designed or operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established by the rules contained in 
the ASME Section III Code. Any future changes to the cooldown or 
heatup rates will be evaluated using 10 CFR 50.59, ``Changes, Tests 
and Experiments,'' and are required to meet the ASME Code margins.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of application for amendment: February 12, 2004.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) Table 3.3.6.1-1, ``Primary 
Containment Isolation Instrumentation,'' to increase the TS Allowable 
Value (AV) related to the setpoint for the Main Steam Tunnel 
Temperature--High system isolation function for those instruments 
located within the Reactor Building. A new Function, 1.f, would be 
added to represent the Reactor Building Main Steam Tunnel Temperature--
High. Existing Function 1.e would be renamed to clarify that it 
represents only the Turbine Building Main Steam Tunnel Temperature--
High.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The leak detection instrumentation associated with the proposed 
amendment is designed to detect Main Steam Line leakage in the range 
of one to ten percent of rated steam flow. This design basis remains 
unchanged. This ensures that the criteria for acceptance as 
established in the original licensing bases remains valid. The 
previous analysis for establishing the allowable value for Main 
Steam Line Tunnel High temperature in the Reactor Building can be 
improved using industry standard, state of the art computer modeling 
techniques. The new analysis using the GOTHIC computer code is 
appropriate because it accurately accounts for the building heat 
structures, HVAC effects, and outside air temperatures. The proposed 
change increases the operating margin, which reduces the potential 
for unnecessary plant transients. Raising the setpoint causes a 
greater time to detect the leak, but remains bounded by existing 
analysis for the design basis break of the main steam line 
documented in Table 14.9.8 of the Peach Bottom [Updated Final Safety 
Analysis Report] UFSAR. There are no impacts on equipment 
qualification. Changes to the instrumentation used to detect a steam 
line leak do not affect the probability of occurrence of the leak. 
Hence, it is concluded that raising the allowable value for Reactor 
Building Main Steam Tunnel high temperature does not significantly 
increase the probability or consequences of an accident previously 
evaluated.

[[Page 9861]]

    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 amendment does not impact the physical design or 
location of the associated leak detection instrumentation. The leak 
detection instrumentation associated with the proposed amendment 
will continue to detect main steam line leakage in the range of one 
to ten percent of rated steam flow. The instruments will still 
initiate the automatic isolation of the appropriate containment 
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 leak in the Reactor Building 
portion of the Main Steam Tunnel and has no impact on the initiation 
of this leak. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Steam leaks in the affected area of the Reactor Building will be 
detected on a timely basis so that the Group 1 Primary Containment 
Isolation Valves are promptly closed. The analysis performed for the 
proposed amendment demonstrates that the appropriate instruments 
will promptly initiate automatic system isolation upon sensing a 
temperature in excess of the new setpoint. Therefore, the proposed 
amendment ensures that the criteria for acceptance as established in 
the original licensing bases remain valid. Further, the proposed 
amendment eliminates a potential cause for unnecessary plant 
shutdowns created by conditions other than a main steam line leak. 
Equipment qualification and structural integrity of systems, 
structures, and components located within the Reactor 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: Mr. Edward Cullen, Vice President and 
General Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
S23-1, Philadelphia, PA 19101.
    NRC Acting Section Chief: Darrell J. Roberts.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: February 4, 2004.
    Description of amendment request: This amendment request proposes 
to update the Technical Specifications (TSs) to correct a non-
conservatism in a TS Table, correct a reference error, update titles, 
incorporate formatting changes to increase ease of use, and remove a 
permit issuance date to ease administrative burden.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Section 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 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 editorial in nature and only correct, update 
and modify the Technical Specifications and Environmental Protection 
Plan.
    The proposed changes do not affect the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated in the 
Seabrook Station UFSAR. Further, the proposed changes do not 
increase the types and amounts of radioactive effluent that may be 
released offsite, and do not significantly increase individual or 
cumulative occupational/public radiation exposures.
    Based on the above, the proposed changes will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes 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 editorial in nature and only update 
Seabrook Station Technical Specifications and Environmental 
Protection Plan to provide consistency and facilitate ease of use. 
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 do not affect plant safety. The plant's design and design 
basis are not revised and the current safety analyses remain in 
effect.
    Thus, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The propose changes do not involve a significant reduction in 
the margin of safety.
    The proposed changes are editorial changes to the Seabrook 
Station Technical Specifications and Environmental Protection Plan. 
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 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: M.S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    Acting NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: December 9, 2003.
    Description of amendment request: The proposed amendment request 
would: (1) Incorporate into the Updated Safety Analysis Report the 
overall Main Steam Isolation Valve (MSIV) Leakage Pathway configuration 
(including the post-accident manual actions necessary to establish that 
configuration) upon Nuclear Regulatory Commission (NRC) approval, (2) 
incorporate into the Cooper Nuclear Station (CNS) licensing basis the 
loss-of-coolant accident (LOCA) dose calculation methodology (currently 
approved on an interim basis) upon permanent approval by the NRC, and 
(3) delete License Condition 2.C.(6), eliminating the commitment to 
provide potassium iodide to the control room occupants during LOCA 
conditions with core damage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 9862]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    The ALT [alternate leakage treatment] pathway was determined 
using the NRC-endorsed method described in Reference 7.3 [NEDC-
31858P-A Class III, August 1999, ``BWROG [Boiling Water Reactor 
Owners Group] Report for Increasing MSIV Leakage Rate Limits and 
Elimination of Leakage Control Systems'']. The proposed manual 
actions to establish that configuration are designed to assure that 
MSIV leakage resulting after a LOCA with core damage will reach the 
Main Turbine Condenser via a pathway that has been evaluated as 
being seismically robust. The LOCA dose calculation methodology 
assumes this leakage reaches the turbine condenser complex. The 
manual actions are simple to perform and there are no concerns for 
personnel safety in carrying out these actions within the timeframes 
established. Accordingly, there is no significant increase in 
probability or consequences of a previously evaluated accident.
    The LOCA dose calculation methodology is already approved on an 
interim basis, as documented in Reference 7.1 [letter to C. Warren 
(NPPD) [Nuclear Public Power District] from U.S. Nuclear Regulatory 
Commission dated February 21, 2003, ``Cooper Nuclear Station--
Issuance of Amendment Regarding Design Basis Accidents'' 
Radiological Dose Assessment Methodologies, and Revision to License 
Condition 2.C.(6) (TAC No. MB4654)'']. As there are no technical 
issues to resolve, the effects of permanent approval on the 
probability or consequences of an accident are bounded by the 
previous safety conclusions of License Amendment 196.
    The deletion of License Condition 2.C.(6), following 
implementation of the seismic evaluation and permanent approval of 
the LOCA dose calculation methodology, is an administrative change 
to the CNS Operating License. Therefore, there are no associated 
effects on the probability or consequences of previously evaluated 
accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    The proposed changes only involve the treatment of the Loss-of-
Coolant Accident. No other new or different kinds of accidents can 
be created by the proposed changes.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    The LOCA dose calculation methodology credits MSIV leakage 
plateout in the Main Turbine Condenser prior to release to the 
Turbine Building. The ALT pathway to the Main Turbine Condenser was 
determined using the NRC-endorsed method described in Reference 7.3. 
Therefore, the effects on safety margins due to crediting this 
configuration are bounded by the NRC Safety Evaluation conclusions 
on this methodology. Using the MSIV leakage assumed in the LOCA 
analysis and conservative assumptions, there is sufficient time for 
the CNS personnel to take the simple actions necessary to configure 
the pathway, and thereby assure that the radiological consequences 
are bounded by the LOCA dose calculation methodology results. 
Accordingly, there is no significant reduction in safety margin.
    The LOCA dose calculation methodology is already approved on an 
interim basis, as documented in Reference 7.1. As there are no 
technical issues to resolve, the effects of permanent approval on 
the [ ] [margin of safety] are bounded by the previous safety 
conclusions of License Amendment 196.
    The deletion of License Condition 2.C.(6), following 
implementation of the seismic evaluation and permanent approval of 
the LOCA dose calculation methodology, is an administrative change 
to the CNS Operating License. Therefore, there are no associated 
effects on safety margins.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa;Docket No. 50-305, Kewaunee Nuclear Power 
Plant, Kewaunee County, Wisconsin;Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota;Docket No. 50-255, Palisades 
Plant, Van Buren County, Michigan;Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin;Docket Nos. 50-282 and 50-306, Prairie Island Nuclear 
Generating Plant, Units 1 and 2, Goodhue County, Minnesota

    Date of amendment request: January 30, 2004.
    Description of amendment request: The proposed amendment deletes 
requirements in the Technical Specifications (TS) to maintain hydrogen 
recombiners and hydrogen and oxygen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI, Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TS for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated January 30, 2004.
    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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG 1.97 Category 1 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen and oxygen monitors no 
longer meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44, the Commission found

[[Page 9863]]

that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
classification of the oxygen monitors as Category 2, and removal of 
the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs, the 
emergency plan (EP), the emergency operating procedures (EOP), and 
site survey monitoring that support modification of emergency plan 
protective action recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen and oxygen monitor equipment was intended to mitigate a 
design-basis hydrogen release. The hydrogen recombiner and hydrogen 
and oxygen monitor equipment are not considered accident precursors, 
nor does their existence or elimination have any adverse impact on 
the pre-accident state of the reactor core or post accident 
confinement of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors. Category 2 oxygen monitors are adequate to verify the 
status of an inerted containment.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    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: Jonathan Rogoff, Morgan Lewis, 1111 
Pennsylvania Avenue NW.,Washington, DC 20004.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: December 1, 2003.
    Description of amendment request: The proposed changes to the Fort 
Calhoun Technical Specifications (TSs) consist primarily of 
typographical changes and relocation of material not required to be in 
the TSs. The licensee has proposed changes to the following TSs: (1) 
Item 14 of Table 3-3 regarding testing of nuclear detector well cooling 
annulus exit air temperature detectors, (2) the title of Item of 10a.2 
of Table 3-5, (3) TS Section 3.17(5)(ii), (4) TS Section 5.5, ``Review 
and Audit,'' (5) TS Section 5.6, ``Reportable Event Action,'' (6) TS 
Sections 5.7.1.b, 5.7.1.c, and 5.7.1.d, (7) TS Section 5.9.1.a, 
``Startup Report,'' and (8) TS Section 5.9.4.c, ``Fire Protection 
Program Deficiency Report.''
    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 relocates requirements for Nuclear Detector 
Cooling that do not meet the criteria for inclusion in the TS set 
forth in 10 CFR 50.36(c)(2)(ii). The requirements for Nuclear 
Detector Cooling are being relocated from TS to the USAR [Updated 
Safety Analysis Report], which will be maintained pursuant to 10 CFR 
50.59, thereby reducing the level of regulatory control. The level 
of regulatory control has no impact on the probability or 
consequences of an accident previously evaluated. Therefore, the 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The correction of typographical errors and relocation of 
specifications is not an initiator of any previously evaluated 
accident. The proposed changes will not prevent safety systems from 
performing their accident mitigation function as assumed in the 
safety analysis.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change relocates requirements for Nuclear Detector 
Cooling that do not meet the criteria for inclusion in TS set forth 
in 10 CFR 50.36(c)(2)(ii). The proposed change only affects the 
technical specifications and does not involve a physical change to 
the plant. Modifications will not be made to existing components nor 
will any new or different types of equipment be installed. The 
proposed change corrects typographical errors and relocates 
information that is unnecessary in the TS. This change will not 
alter assumptions made in safety analysis and licensing bases.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change relocates requirements for Nuclear Detector 
Cooling that do not meet the criteria for inclusion in TS set forth 
in 10 CFR 50.36(c)(2)(ii). The change will not reduce a margin of 
safety since the location of a requirement has no impact on any 
safety analysis assumptions. In addition, the relocated requirements 
for Nuclear Detector Cooling remain the same as the existing TS. 
Since any future changes to these requirements or the surveillance 
procedures will be evaluated per the requirements of 10 CFR 50.59, 
there will be no reduction in a margin of safety.
    The additional proposed changes correct typographical errors and 
relocate redundant information not required to be in the TS.

[[Page 9864]]

    Therefore, this technical specification change does not involve 
a significant reduction in the margin of safety.

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

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment requests: December 30, 2003.
    Description of amendment requests: The proposed amendment deletes 
the requirements from the technical specifications (TS) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TS for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 30, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    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: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

[[Page 9865]]

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

    Date of application for amendments: December 30, 2003.
    Brief description of amendments: The amendment revised the 
Administrative Controls Section 5.1.5 to state any Senior Reactor 
Operator may be designated to be responsible for the control room 
command function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.92(c), 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?
    The proposed change to Technical Specifications Administrative 
Controls Section 5.1.5, involves the use of a more generic 
designation of SRO [Senior Reactor Operator] for the unit staff 
position responsible for the control room command function. Since 
the proposed change is administrative in nature, it does not involve 
any physical changes to any structures, systems, or components, nor 
will their performance requirements be altered. The proposed change 
also does not affect the operation, maintenance, or testing of the 
plant. Therefore, the response of the plant to previously analyzed 
accidents will not be affected. Consequently, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    As a result of the proposed change to the Technical 
Specifications, the qualification requirements for the unit staff 
position responsible for the control room command function will 
remain unchanged and the plant staff will continue to meet 
applicable regulatory requirements. Also, since no change is being 
made to design, operation, maintenance, or testing of the plant, no 
new methods of operation or failure modes are introduced by the 
proposed change. Therefore, the possibility of a new or different 
kind of accident from any previously evaluated is not created.
    3. Does the proposed change involve a significant decrease in 
the margin of safety?
    The proposed change to the Technical Specifications will have no 
adverse impact on the onsite organizational features necessary to 
assure safe operation of the plant since the qualification 
requirements for the unit staff position for the control room 
command function remain unchanged. The adoption of the more generic 
designation of SRO for the individual responsible for control room 
command function will also reduce the regulatory burden of having to 
devote limited resources to process a license amendment whenever a 
title change for this position is implemented, thus improving plant 
efficiency. Therefore, the proposed change does not invoice a 
significant decrease in the margin of safety.

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

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

    Date of amendment request: February 3, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
4.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
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 April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated February 3, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change 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). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS Limiting 
Conditions for Operation (LCO). The risk associated with this 
allowance is managed by the imposition of required actions that must 
be performed within the prescribed completion times. The net effect 
of being in a TS condition on the margin of safety is not considered 
significant.

[[Page 9866]]

The proposed change does not alter the required actions or 
completion times of the TS. The proposed change allows TS conditions 
to be entered, and the associated required actions and completion 
times to be used in new circumstances. This use is predicated upon 
the licensee's performance of a risk assessment and the management 
of plant risk. The change also eliminates current allowances for 
utilizing required actions and completion times in similar 
circumstances, without assessing and managing risk. The net change 
to the margin of safety is insignificant. 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: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 21, 2004.
    Brief description of amendments: The amendment would revise 
Technical Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' and 3.3.6, ``Containment Ventilation 
Isolation Instrumentation.'' The purpose of the amendment is to adopt 
the completion time, test bypass time, and surveillance frequency time 
changes approved by the NRC in Topical Reports WCAP-14333-P-A, 
``Probabilistic Risk Analysis of the RPS [reactor protection system] 
and ESFAS Test Times and Completion Times,'' and WCAP-15376-P-A, 
``Risk-Informed Assessment of the RTS and ESFAS Surveillance Test 
Intervals and Reactor Trip Breaker Test and Completion Times.'' The 
proposed changes would revise the required actions for certain action 
conditions; increase the completion times for several required actions 
(including some notes); delete notes in certain required actions; and 
increase frequency time intervals (including certain notes) in several 
surveillance requirements (SRs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The same reactor trip system (RTS) 
and engineered safety feature actuation system (ESFAS) 
instrumentation will continue to be used. The protection systems 
will continue to function in a manner consistent with the plant 
design basis. These changes to the Technical Specifications [in the 
amendment] do not result in a condition where the design, material, 
and construction standards that were applicable prior to the change 
are altered.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators [because the proposed changes are not event initiators]. 
There will be no degradation in the performance of or an increase in 
the number of challenges imposed on safety-related equipment assumed 
to function during an accident situation. There will be no change to 
normal plant operating parameters or accident mitigation 
performance. The proposed changes will not alter any assumptions or 
change any mitigation actions in the radiological consequence 
evaluations in the FSAR [Comanche Peak Final Safety Analysis 
Report].
    The determination that the results of the proposed changes are 
acceptable [to be considered for plant-specific Technical 
Specifications] was established in the NRC Safety Evaluations 
prepared for WCAP-14333-P-A (issued by letter dated July 15, 1998) 
and for WCAP-15376-P-A (issued by letter dated December 20, 2002). 
Implementation of the proposed changes will result in an 
insignificant risk impact. Applicability of these conclusions has 
been verified through plant-specific reviews and implementation of 
the generic analysis results in accordance with the respective NRC 
Safety Evaluation conditions [for the two WCAPs].
    The proposed changes to the Completion Times, test bypass times, 
and Surveillance Frequencies reduce the potential for inadvertent 
reactor trips and spurious ESF [engineered safety feature] 
actuations, and therefore do not increase the probability of any 
accident previously evaluated. The proposed changes do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the RTS and ESFAS signals. The RTS and 
ESFAS 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 the increase in core damage frequency (CDF) is less than 
1.0E-06 per year and the increase in large early release frequency 
(LERF) is less than 1.0E-07 per year. In addition, for the 
Completion Time changes, the incremental conditional core damage 
probabilities (ICCDP) and incremental conditional large early 
release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, 
respectively. These changes meet the acceptance criteria in 
Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and 
ESFAS will continue to perform their [safety] functions with high 
reliability as originally assumed, and the increase in risk as 
measured by ``CDF, ``LERF, ICCDP, ICLERP risk metrics is within the 
acceptance criteria of existing [NRC] regulatory guidance, there 
will not be a significant increase in the consequences of any 
accidents.
    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. The proposed changes are consistent with safety analysis 
assumptions and resultant consequences.
    Therefore, [the] change[s do] not increase the probability or 
consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The proposed changes will not affect the normal method of 
plant operation. No performance requirements will be affected or 
eliminated. The proposed changes will not result in physical 
alteration to any plant system nor will there be any change in the 
method by which any safety-related plant system performs its safety 
function.
    There will be no setpoint changes or changes to accident 
analysis assumptions.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. There will be no impact on the overpower limit, DNBR 
[departure from nucleate boiling ratio] limits, FQ [heat 
flux hot channel factor], F[Delta]H [nuclear enthalpy 
rise hot channel factor], LOCA PCT [loss-of-

[[Page 9867]]

coolant accident peak cladding temperature], peak local power 
density, or any other margin of safety. The radiological dose 
consequence acceptance criteria listed in the [NRC] Standard Review 
Plan will continue to be met.Redundant RTS and ESFAS trains are 
maintained, and diversity with regard to the signals that provide 
reactor trip and engineered safety features actuation is also 
maintained. All signals credited as primary or secondary, and all 
operator actions credited in the accident analyses will remain the 
same. The proposed changes will not result in plant operation in a 
configuration outside the design basis. The calculated impact on 
risk is insignificant and meets the acceptance criteria contained in 
Regulatory Guides 1.174 and 1.177. Although there was no attempt to 
quantify any positive human factors benefit due to increased 
Completion Times and bypass test times, it is expected that there 
would be a net benefit due to a reduced potential for spurious 
reactor trips and actuations associated with testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
    (a) Reduced testing will result in fewer inadvertent reactor 
trips, less frequent actuation of ESFAS components, less frequent 
distraction of operations personnel without significantly affecting 
RTS and ESFAS reliability.
    (b) Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation will be realized. This is 
due to less frequent distraction of the operators and shift 
supervisor to attend to instrumentation Required Actions with short 
Completion Times.
    (c) Longer repair times associated with increased Completion 
Times will lead to higher quality repairs and improved reliability.
    (d) The Completion Time extensions for the reactor trip breakers 
will provide the utilities additional time to complete test and 
maintenance activities while at power, potentially reducing the 
number of forced outages related to compliance with reactor trip 
breaker Completion Times, and provide consistency with the 
Completion Times for the logic trains.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

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

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed NoSignificant Hazards 
Consideration Determination, and Opportunity for a Hearing

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

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3

    Date of amendment request: November 13, 2003.
    Brief description of amendment request: The proposed amendment 
would allow an increase in the licensed power from 3441 megawatts 
thermal (MWt) to 3716 MWt. This represents an increase of approximately 
8 percent above the current rated licensed thermal power. The proposed 
amendment would also change the operating license and the technical 
specifications appended to the operating license to provide for 
implementing uprated power operation.
    Date of publication of individual notice in Federal Register: 
February 5, 2004.
    Expiration date of individual notice: March 8, 2004.

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

    Date of amendment request: January 30, 2004.
    Brief description of amendment request: The proposed amendment 
would revise the Cooper Nuclear Station (CNS) Technical Specifications 
(TS), by adding a temporary note to allow a one-time extension of a 
limited number of TS Surveillance Requirements (SRs). The temporary 
note states that the next required performance of the SR may be delayed 
until the current cycle refueling outage, but no later than February 2, 
2005, and it expires upon startup from the refueling outage. With the 
exception of one SR, the period of additional time requested occurs 
during the next planned refueling outage.
    Date of publication of individual notice in Federal Register: 
February 12, 2004 (69 FR 7023).
    Expiration date of individual notice: March 15, 2004.

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 email to pdr@nrc.gov.

[[Page 9868]]

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

    Date of application for amendment: December 2, 2003.
    Brief description of amendment: The amendment revised Surveillance 
Requirement (SR) 4.0.2 of the Technical Specifications (TSs) to extend 
the delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance. The delay period is extended from the 
current limit of ``* * * up to 24 hours or up to the limit of the 
specified frequency, whichever is less'' to ``* * * up to 24 hours or 
up to the limit of the specified frequency, whichever is greater.'' The 
revised SR 4.0.2 specifies that a risk evaluation shall be performed 
for any surveillance delayed greater than 24 hours and the risk impact 
shall be managed. In addition, a new Section 6.21 is added to provide 
for a TS Bases Control Program.
    Date of Issuance: February 5, 2004.
    Effective date: February 5, 2004 and shall be implemented within 60 
days of issuance.
    Amendment No.: 240.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
692).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated February 5, 2004.
    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: June 11, 2003, as supplemented 
August 20 and October 13, 2003.
    Brief description of amendment: The amendment allows the licensee 
to extend its Appendix J, Type A, Containment Integrated Leak Rate 
Test, Option B, for H. B. Robinson Steam Electric Plant, Unit No. 2, 
from the scheduled May 2004 timeframe to no later than April 9, 2007.
    Date of issuance: February 11, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No. 199.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 23, 2003 (68 
FR 74264).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 11, 2004.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket No. 50-270, Oconee Nuclear Station, 
Unit 2, Oconee County, South Carolina

    Date of application of amendment: October 28, 2003.
    Brief description of amendment: The amendment revised the licensing 
basis in the Updated Final Safety Analysis Report (UFSAR) to support 
installation of a passive low-pressure injection (LPI) cross connect 
inside containment. The changes to the UFSAR revise the licensing basis 
for selected portions of the core flood and LPI/Decay Heat Removal 
piping to allow exclusion of the dynamic effects associated with 
postulated rupture of that piping by application of leak-before-break 
technology.
    Date of Issuance: February 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 338.
    Renewed Facility Operating License No. DPR-47: Amendment revised 
the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68661)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 5, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 12, 2003, as revised by 
letters dated December 5 and 18, 2003.
    Brief description of amendment: By letter dated December 5, 2003, 
Entergy submitted a revised application for amendment to Grand Gulf 
Nuclear Station, Unit 1 Technical Specification (TS) 3.3.6.1, ``Primary 
Containment and Drywell Isolation Instrumentation,'' to add a provision 
to the APPLICABILITY function that will eliminate the requirement that 
the Residual Heat Removal System Isolation, Reactor Vessel Water Level-
Low, Level 3, be OPERABLE under certain conditions during refueling 
outages. Specifically, the proposed change requested in the original 
application dated May 12, 2003, would remove the requirement for this 
isolation function, specified in Table 3.3.6.1-1, when the upper 
containment reactor cavity is at the High Water Level condition 
specified in TS 3.5.2, ``Emergency Core Cooling Systems (ECCS) 
Shutdown.'' The revised application adds a new surveillance requirement 
(SR) (SR 3.3.6.1.9) to verify every four hours that the water level in 
the upper containment pool is greater than or equal to 22 feet 8 inches 
above the reactor pressure vessel flange, and adds a footnote to Table 
3.3.6.1-1, Item 5.b, for MODE 5 that states that the function is not 
required when the upper containment reactor cavity and transfer canal 
gates are removed and SR 3.3.6.1.9 is met. The proposed SR and footnote 
are only applicable in MODE 5. The May 12, 2003, application was 
previously noticed in the Federal Register on June 10, 2003 (68 FR 
34665).
    Date of issuance: January 23, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 163.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 2003 (68 
FR 69726). The December 18, 2003, supplemental letter provided 
clarifying information that did not change the scope of the December 
15, 2003, Federal Register notice or the no significant hazards 
consideration determination therein.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 23, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 15, 2003, as supplemented 
by letter on September 15, 2003.
    Brief description of amendment: The amendment revised the reactor 
coolant system pressure-temperature limit curves in Section 3.4.11, 
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,'' 
of the Technical Specifications. The revised curves are effective up to 
22 effective full-power years.
    Date of issuance: January 27, 2004.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 110.

[[Page 9869]]

    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 2, 2003 (68 
FR 52235).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated January 27, 2004.
    No significant hazards consideration comments received: No
    The September 15, 2003, letter provided clarifying information 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination.

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

    Date of application for amendment: August 28, 2003.
    Brief description of amendment: The amendment revised Section 
3.1.7, ``Standby Liquid Control (SLC) System,'' of the Technical 
Specifications to support a transition from GE11 to GE14 fuel in the 
reactor core. The revised Section 3.1.7 raises the required calculated 
average boron concentration in the reactor from a concentration 
equivalent to 660 parts per million (ppm) natural boron to 780 ppm 
natural boron. The increased concentration is achieved by requiring use 
of sodium pentaborate solution enriched with the boron-10 isotope.
    Date of issuance: February 13, 2004.
    Effective date: As of the date of issuance to be implemented prior 
to startup from Refueling Outage 9.
    Amendment No.: 111.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 30, 2003 (68 
FR 56345). The staff's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 14, 2003, as 
supplemented on October 2, 2003.
    Brief description of amendments: The amendments modify the Salem 
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications 
(TSs) by: (1) Adding new TS 3/4.7.11, ``Fuel Storage Pool Boron 
Concentration,'' to define spent fuel pool boron concentration limits; 
(2) relocating fuel assembly storage requirements currently located in 
TS 5.6.1.2d to a new TS 3/4.7.12, ``Fuel Assembly Storage in the Spent 
Fuel Pool;'' and (3) relocating refueling boron concentration 
requirements from TS 3/4.9.1, ``Boron Concentration,'' to the Core 
Operating Limits Report.
    Date of issuance: February 6, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 262 and 244.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: April 29, 2003 (68 FR 
22753).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 6, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 13, 2002, as supplemented 
on April 1 and November 21, 2003.
    Brief description of amendment: The amendment approves revisions to 
the Updated Final Safety Analysis Report (UFSAR) to update the quality 
assurance criteria and the basis for the seismic qualification of the 
ducting installed as part of the suspended ceiling air delivery system 
in the main control room.
    Date of issuance: February 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
in accordance with 10 CFR 50.71(e).
    Amendment No.: 50.
    Facility Operating License No. NPF-90: Amendment revised the UFSAR.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18286). The supplemental letters provided clarifying information that 
did not expand the scope of the original request and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 12, 2004.
    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 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

[[Page 9870]]

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 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 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. Within 60 days after the date 
of publication of this notice, 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 2.309, 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 a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
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. The 
petition must include 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/requestor 
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. 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 by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. 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 it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by email to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and

[[Page 9871]]

petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

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

    Date of amendment request: February 6, 2004.
    Description of amendment request: The amendment changes the 
implementation date from 30 days to 120 days for Amendment No. 224 
issued on January 16, 2004, that approved a measurement uncertainty 
uprate to increase the licensed rated power by 1.6 percent from 1500 
megawatts thermal (MWt) to 1524 MWt.
    Date of issuance: February 13, 2004.
    Effective date: February 13, 2004, and the fully implemented date 
for Amendment No. 224 (issued January 16, 2004) is changed to 120 days.
    Amendment No.: 225.
    Renewed Facility Operating License No. DPR-40: Amendment revises 
the implementation date for Amendment No. 224.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Omaha-World Herald. The notice provided an 
opportunity to submit comments on the Commission's proposed NSHC 
determination. No comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, State consultation, and final NSHC determination 
are contained in a safety evaluation dated February 13, 2004.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of application for amendment: February 5, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 3.7.5, ``Auxiliary Feedwater (AFW) System'' to 
incorporate a one-time provision that extends the allowed outage time 
for an inoperable turbine-driven auxiliary feedwater pump.
    Date of issuance: February 6, 2004.
    Effective date: February 6, 2004.
    Amendment No.: 158.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed 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 
February 6, 2004.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

    Dated at Rockville, Maryland, this 20th day of February 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management,Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-4343 Filed 3-1-04; 8:45 am]
BILLING CODE 7590-01-P