Operating licenses, amendments; no significant hazards considerations; biweekly notices,

[Federal Register: August 30, 2005 (Volume 70, Number 167)]

[Notices]

[Page 51378-51388]

From the Federal Register Online via GPO Access [wais.access.gpo.gov]

[DOCID:fr30au05-71]

NUCLEAR REGULATORY COMMISSION

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

1. 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 August 5, 2005, to August 18, 2005. The last biweekly notice was published on August 16, 2005 (70 FR 48201).

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. 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 within 60 days, 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

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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 basis 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 satisfy 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. 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 e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition

for leave to intervene should also be sent to the attorney for the licensee.

Nontimely 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 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: July 14, 2005.

Description of amendment request: The proposed change would modify the Millstone Power Station, Unit No. 2 reactor coolant system (RCS) heatup and cooldown limits Technical Specification (TS) 3.4.9.1, ``Reactor Coolant System''. The associated TS bases will be updated to address the proposed changes.

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

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

   Response: No.

   The proposed changes are a result of the new analysis of the RCS P-T [pressure-temperature] limits and associated heatup/cooldown rates. These changes will support plant operation to 54 EFPY

   [effective full-power years] and provide flexibility during plant heatup and cooldown, especially during equipment manipulations such as securing RCPs [reactor coolant pumps], swapping SDC [shutdown cooling] heat exchangers, and initiating SDC.

   The hydrostatic and leak test limit will now be administratively controlled by the heatup limit. Administratively limiting hydrostatic and leak tests to the heatup limit provides additional margin to the Appendix G requirements. Table 3.4-2 has been modified to remove the Inservice Hydrostatic and Leak Testing item and to add a note indicating heatup limitations also apply to hydrostatic and leak test conditions. The requirement to remain isothermal (rate `` 5 [deg]F/hour) for 1 hour prior [to] and during hydrostatic and leak test [s] above the heatup curve is no longer needed as operation above the heatup curve is no longer allowed.

   The proposed changes to the RCS P-T limits and rates of temperature change are based on the new analysis. This analysis uses standard approved methods that ensure the margins of safety required by 10 CFR 50, Appendix G are maintained. The other changes discussed are more restrictive enhancements to technical specification requirements. Therefore, the proposed changes will not result in a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

   Response: No.

   The proposed changes will not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. They do not alter the way any structure, system, or component functions. The increased heatup and cooldown rates are bounded by the existing accident analysis. The proposed changes do not introduce any new failure modes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

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   Response: No.

   The proposed changes will modify the RCS P-T limits, and the RCS heatup and cooldown rate limits. The proposed changes are being made as a result of the new P-T and LTOP [low-temperature overpressure protection] analyses performed. The new P-T curves and heatup and cooldown rates are developed in accordance with the requirements and methods described in 10 CFR 50 Appendix G and are consistent with the criteria contained in the Standard Review Plan Section 5.3.2. This will ensure the integrity of the reactor vessel is maintained during all aspects of plant operation. Therefore, there is no significant effect on the probability or consequences of any accident previously evaluated and no significant impact on offsite doses associated with previously evaluated accidents. This license amendment request does not result in a reduction of the margin of safety as defined in the bases for the technical specifications addressed by the proposed changes.

   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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

   NRC Section Chief: Darrell J. Roberts.

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

   Date of amendment request: May 24, 2005.

   Description of amendment request: The proposed amendment would revise the Technical Specification allowances to bypass the rod worth minimizer consistent with previously-approved standards.

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

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

   Response: No. The proposed special operation allowances do not involve the modification of any plant equipment or affect basic plant operation. The relevant design basis accident is the control rod drop accident (CRDA), which involves multiple failures to initiate the event. Control rod decoupling and remaining stuck full- in while its drive mechanism is withdrawn are required initiators. The proposed special operations have no adverse impact on control rod coupling or control rod performance. As such, there is no significant increase in the probability of an accident previously evaluated.

   The CRDA analysis consequences and related initial conditions remain unchanged when invoking the proposed special operation allowance. The control rod withdrawal sequence is assumed to limit individual control rod worths as another initial condition for the CRDA. However, consistent with existing requirements for control rod withdrawal operations, all control rod withdrawal sequences are analyzed to meet this criterion and are implemented under the control of the rod worth minimizer or by independent verification by a second licensed operator or other qualified member of the technical staff. The consequences of analyzed events are therefore not affected. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

   Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

   Response: No. The proposed special operation allowances do not involve the modification of any plant equipment or affect basic plant operation. The relevant design basis accident is the control rod drop accident (CRDA), which involves multiple failures to initiate the event. Additionally, CRDA analysis consequences and related initial conditions remain unchanged when invoking the proposed special operation allowance. These changes do not negate any existing requirement, and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. 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: J.M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 02360-5599.

   NRC Section Chief: Darrell Roberts.

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

   Date of amendment request: May 24, 2005.

   Description of amendment request: The proposed amendment would revise Technical Specification (TS) applicability requirements related to primary containment oxygen concentration and drywell-to-suppression chamber differential pressure limits.

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

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

   Response: No. The proposed applicability and associated default actions being revised do not involve the modification of any plant equipment or affect basic plant operation. Additionally, the associated limitations are not assumed to be an initiator of any analyzed event. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

   The limits imposed by the associated specifications remain unchanged. The consequences of analyzed events are therefore not affected. Brief periods where the requirements for maintaining these limits are relaxed are currently considered in the TS and associated licensing basis. The proposed change clarifies and modifies the definition of these periods, however, any changes are not considered significant and are supported by remaining [definitions] consistent with the recommended allowances of NUREG-1433, Rev. 3, ``Standard Technical Specifications, General Electric Plants, BWR [boiling- water reactor]/4.'' Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

   Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with

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   current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

   Response: No. The proposed applicability and associated default actions being revised do not involve the modification of any plant equipment or affect basic plant operation. Additionally, the associated limitations remain unchanged. These changes do not negate any existing requirement, and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. The revised plant conditions reflecting the applicability and the duration allowed to restore limits are not credited in any design basis event. These changes do not reflect any significant adverse impact to the overall risk of operating during brief periods without the required primary containment oxygen concentration since the total time for any occurrence is only marginally extended and reflects times recommended by NUREG-1433. 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: J.M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 02360-5599.

   NRC Section Chief: Darrell Roberts.

   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 amendment request: June 27, 2005.

   Description of amendment request: The proposed amendment revises the Facilities Operating License to change technical specification (TS) 3.6.1.3, Required Actions A.1 and B.1, to add closed relief valves as acceptable isolation devices provided that the relief setpoint is greater than 1.5 times containment design pressure.

   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.

   Primary Containment Isolation Valves (PCIVs) are accident mitigating features designed to limit releases from the containment following an accident. The Technical Specifications (TS) specify actions to be taken to preserve the containment isolation function if a PCIV is inoperable. These actions include isolating the penetration flow path by specific methods. The proposed TS change adds closed relief valves with acceptable relief setpoints as another method to isolate the penetration flowpath. The use of relief valves with relief setpoints greater than 1.5 times the containment design pressure meets the Standard Review Plan options for acceptable isolation devices. This relief setpoint provides [a] sufficient margin to minimize the potential for premature opening due to containment post-accident pressures. The proposed change does not affect any initiators to accidents previously evaluated.

   Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

   Response: No.

   The proposed change does not introduce any new modes of plant operation or adversely affect the design function or operation of safety features. The proposed TS change allows use of existing plant equipment as compensatory measures to maintain the containment isolation design intent when the normal isolation features are inoperable. Since relief valves used for this purpose will not be disabled by blind flanges, the system piping overpressure protection design feature will also be preserved.

   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 safety margin associated with this change is that associated with preserving the containment integrity. NUREG-0800, the Standard Review Plan, recognizes that relief valves with relief setpoints greater than 1.5 times containment design pressure are acceptable as containment isolation devices. Closed relief valves with relief setpoints of this margin provide an isolation alternative that is less susceptible to a single failure (i.e., inadvertent opening) yet still preserves the overpressure protection that the component was intended to provide.

   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, 1700 K Street, NW., Washington, DC 20006-3817.

   NRC Section Chief: David Terao.

   Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

   Date of amendment request: June 15, 2005.

   Description of amendment request: The proposed change revises Technical Specification (TS) 3.3.2.2 ``Feedwater System and Main Turbine High Water Level Trip Instrumentation,'' to reflect a design change to the instrumentation logic.

   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?

   The proposed change revises TS 3.3.2.2 to reflect a design change to the instrumentation logic that trips the three feedwater pumps and main turbine. The design change will add a redundant high reactor water level trip channel to both trip systems. The Feedwater System and main turbine high water level trip is credited in the QCNPS [Quad Cities Nuclear Power Station] accident analysis to function during an increase in feedwater flow transient. Specifically, the instrumentation and associated trip limits the reactor water level increase resulting from a feedwater controller failure during maximum flow demand, thus preventing a nuclear fuel minimum critical power ratio violation associated with increased subcooling and resultant pressure transient. Additionally, this trip function prevents excessive water inventory from entering the main steam system and damaging steam-handling equipment.

   TS requirements that govern operability or routine testing of plant instruments are not assumed to be initiators of any analyzed event because these instruments are intended to detect, prevent, or mitigate accidents. The Feedwater System and main turbine trip instrumentation serves to mitigate transients that result in increased reactor water level. The trip instrumentation associated with the proposed changes and design change are independent from the instrumentation and logic used in the Feedwater Control System and Turbine Control System. Therefore, the proposed change does not involve a

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   significant increase in the probability of an accident previously evaluated.

   The proposed design change to add a redundant high reactor water level trip channel to both trip systems, and the associated TS changes, do not adversely impact the instrumentation's ability to perform the functions described above. The design change will utilize installed spare trip units and relay contacts of the same design as those presently credited to meet TS 3.3.2.2 requirements. The method in which the reactor water level is sensed and the reactor water level setpoints at which a trip is initiated are not impacted. The instrumentation response times and the instrumentation output to the equipment being tripped remains the same. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Furthermore, there will be no change in the types or significant increase in the amounts of any effluents released offsite.

   Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

   The proposed change does not alter the parameters within which the plant is operated. There are no setpoints at which protective or mitigative actions are initiated that are affected by the proposed change. This proposed change will not alter the manner in which equipment operation is initiated nor will the demands on mitigating equipment be changed. The proposed change to TS 3.3.2.2 adds redundant instrumentation to improve system reliability, and increase maintenance and testing flexibility. The instrumentation being added to the trip logic utilizes the same transmitters, and the same type of trip units and trip relays, as presently used to monitor reactor water level and initiate Emergency Core Cooling System operation.

   Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

   Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the establishment of setpoints to initiate alarms or actions. The proposed amendment supports a change to the logic that trips the three feedwater pumps and the main turbine from a two-out-of-two initiation logic to a one-out-of-two twice initiation logic. The proposed amendment does not alter the setpoints at which the trip function occurs, the response time of the trip initiation logic, or the plant response following a valid trip signal. The proposed changes to the TS 3.3.2.2 Required Actions and Completion Times are consistent with other instrumentation TS that incorporate a one-out-of-two twice initiation logic.

   Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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

   Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

   NRC Section Chief: Gene Y. Suh.

   PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

   Date of amendment request: August 4, 2005.

   Description of amendment request: The proposed amendment would revise Hope Creek Generating Station Technical Specification 3.7.1.3, ``Ultimate Heat Sink,'' to allow a 24-hour average temperature to be used if ultimate heat sink temperature exceeds 89.5 [deg]F provided the ultimate heat sink temperature or safety auxiliary cooling system temperature does not exceed 95 [deg]F.

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

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

   Response: No.

   The ultimate heat sink (UHS) is not an accident indicator. An increase in UHS temperature will not increase the probability of occurrence of an accident. The proposed change will allow plant operation to continue if temperature of the UHS exceeds 89.5 [deg]F provided that UHS temperature averaged over the previous 24-hour period is less than 89.5 [deg]F and the UHS temperature and safety auxiliary cooling system (SACS) temperatures do not exceed 95

   [deg] F. Maintaining these temperatures less than or equal to 95

   [deg] F ensures that accident mitigation equipment will continue to perform its required function.

   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 of different kind of accident from any accident previously evaluated?

   Response: No.

   The proposed change will not install any new or different equipment or modify equipment in the plant. The proposed change will not alter the operation or function of structures, systems or components. The response of the plant and the operators following a design basis accident is unaffected by this change. The proposed change does not introduce any new failure modes and the design basis heat removal capability of the safety related components is maintained at the increased UHS temperature limit.

   Therefore, the proposed chage will 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 the margin of safety?

   Response: No.

   The increase to the UHS temperature will not adversely affect design basis accident mitigation equipment. Ensuring that SACS temperature remains below 95 [deg]F when UHS is above 89.5 [deg]F ensures that heat removal capability is within the current analyzed limits. Accident mitigation equipment will continue to function as assumed in the accident analysis. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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

   Attorney for Licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

   NRC Section Chief: Darrell J. Roberts.

   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

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   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 (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 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

   AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

   Date of application for amendment: December 17, 2004.

   Brief description of amendment: The amendment revised Appendix B, Environmental Protection Plan (non-radiological) of the Clinton Facility Operating License.

   Date of issuance: August 9, 2005.

   Effective date: As of the date of issuance and shall be implemented within 60 days.

   Amendment No.: 166.

   Facility Operating License No. NPF-62: The amendment revised the Environmental Protection Plan.

   Date of initial notice in Federal Register: April 12, 2005 (70 FR 19112).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 9, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: October 20, 2004.

   Brief description of amendment: The amendment revised Table 4.1-1, ``Instrument Surveillance Requirements,'' of the Technical Specifications and associated Bases to extend the functional testing surveillance interval from monthly to a semi-annual interval for reactor trip system instrumentation channels, and from the current monthly to quarterly for the reactor trip devices.

   Date of issuance: August 11, 2005.

   Effective date: As of the date of issuance and shall be implemented within 60 days.

   Amendment No.: 255.

   Facility Operating License No. DPR-50. Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: November 23, 2004 (69 FR 68181)

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 11, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: June 24, 2004, as supplemented February 24, 2005.

   Brief description of amendment: The amendment revised TMI-1 Technical Specification (TS) 4.0.2 to adopt the provisions of Technical Specification Task Force (TSTF) Traveller TSTF-358, Revision 6, revising the required actions and time constraints regarding missed surveillances. The amendment also added a new Section 6.18 to the TSs incorporating a Technical Specifications Bases Control Program.

   Date of issuance: August 12, 2005.

   Effective date: As of the date of issuance and shall be implemented within 60 days.

   Amendment No.: 256.

   Facility Operating License No. DPR-50. Amendment revised the TSs.

   Date of initial notice in Federal Register: March 1, 2005 (70 FR 9987).

   The supplement dated February 24, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination . The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 12, 2005.

   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 21, 2004.

   Brief description of amendment: The amendment changes Technical Specification Section 5.5.14, ``Technical Specifications (TS) Bases Control Program,'' to incorporate changes in Section 50.59 of Title 10 of the Code of Federal Regulations terminology. The amendment also revises Section 5.7.1, ``High Radiation Area,'' by adding wording that was inadvertently deleted with the issuance of the Improved Standard Technical Specifications in Amendment No. 176.

   Date of issuance: August 2, 2005.

   Effective date: August 2, 2005.

   Amendment No.: 205.

   Renewed Facility Operating License No. DPR-23. Amendment revises the Technical Specifications.

   Date of initial notice in Federal Register: August 31, 2004 (69 FR 53101).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 2, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: December 23, 2004.

   Brief description of amendment: The amendment relocated certain Technical Specifications (TSs) to the Millstone Power Station, Unit No. 3 Technical Requirements Manual.

   Date of issuance: August 11, 2005.

   Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

   Amendment No.: 225.

   Facility Operating License No. NPF-49: The amendment revised the TSs.

   Date of initial notice in Federal Register: May 24, 2005 (70 FR 29788).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 11, 2005.

   No significant hazards consideration comments received: No.

   Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

   Date of amendment request: September 30, 2004, as supplemented by letters dated April 26 and June 8, 2005.

   [[Page 51384]]

   Brief description of amendment: The amendment changes the existing steam generator (SG) tube surveillance program to be consistent with that being proposed by the Technical Specification (TS) Task Force (TSTF) in TSTF-449. These changes revise definitions in TS 1.1, reactor coolant system operational leakage in TS 3.4.13, SG program in TS 5.5.9, and SG tube inspection reports in TS 5.6.7, and add a new TS 3.4.16 on SG tube integrity. Also, as a result of the licensee replacing the SGs with SGs having a new Alloy 690 thermally treated tubing design, the TSs are revised to reflect this replacement.

   Date of issuance: August 10, 2005.

   Effective date: As of the date of issuance and shall be implemented prior to resumption of operation from the 1R19 refueling outage scheduled for the fall of 2005.

   Amendment No.: 224.

   Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications.

   Date of notices in Federal Register: November 9, 2004 (69 FR 64987) and May 24, 2005 (70 FR 29790). The supplement dated June 8, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 10, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: December 6, 2004, as supplemented on June 14, 2005.

   Brief description of amendment: The amendment makes administrative and other miscellaneous changes to the Facility Operating License (FOL) and Technical Specifications (TSs) including correction of references and deleting obsolete or redundant TS requirements and surveillances.

   Date of Issuance: August 15, 2005.

   Effective date: As of the date of issuance, and shall be implemented within 30 days.

   Amendment No.: 226.

   Facility Operating License No. DPR-28: The amendment revised the FOL and TSs.

   Date of initial notice in Federal Register: January 18, 2005 (70 FR 2888). The supplement contained clarifying information only, and did not change the initial no significant hazards consideration determination or expand the scope of the initial Federal Register notice.

   The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated August 15, 2005.

   No significant hazards consideration comments received: No.

   Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50- 277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania.

   Date of application for amendments: September 15, 2004.

   Brief description of amendments: The amendments deleted the Technical Specification requirements to maintain hydrogen recombiners and hydrogen/oxygen monitors and related Surveillance Requirements.

   Date of issuance: August 11, 2005.

   Effective date: As of the date of issuance, to be implemented within 120 days.

   Amendments Nos.: 256 and 259.

   Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: February 1, 2005, (70 FR 5244). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 11, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: April 22, 2005.

   Brief description of amendment: The amendment revised the Technical Specifications (TSs) related to fuel handling and storage. Specifically, the changes revised TS 3/4.9.11, ``Storage Pool Water Level,'' TS 3/4.9.12, ``Storage Pool Ventilation,'' TS 3/4.9.13, ``Spent Fuel Assembly Storage,'' and TS 5.6, ``Fuel Storage,'' to reflect that spent fuel storage racks are no longer installed in the cask pit or transfer pit. Fuel storage racks were permitted to be temporarily installed in the cask pit and transfer pit during a project to increase spent fuel pool (SFP) storage capacity. All temporarily installed fuel storage racks have now been moved into the SFP. Additionally, the changes relocated the requirements of TS 3/4.9.7, ``Crane Travel--Fuel Handling Building,'' to the Davis-Besse Nuclear Power Station Technical Requirements Manual. The changes to TS 3/4.9.13 and TS 5.6 also reflected that there are no longer low density fuel storage racks in the SFP. The changes made TS requirements consistent with the current fuel storage design.

   Date of issuance: August 16, 2005.

   Effective date: As of the date of issuance and shall be implemented within 120 days.

   Amendment No.: 266.

   Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: May 24, 2005 (70 FR 29795).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 16, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: February 22, 2005.

   Brief description of amendment: The amendment revised the Technical Specifications by eliminating the requirements to provide the NRC monthly operating reports and annual occupational radiation exposure reports.

   Date of issuance: August 16, 2005.

   Effective date: As of the date of issuance and shall be implemented within 60 days.

   Amendment No.: 267.

   Facility Operating License No. NPF-3: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: May 10, 2005 (70 FR 24651).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 16, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: February 22, 2005.

   Brief description of amendment: The amendment revised the Technical Specifications by eliminating the requirements to submit the NRC monthly operating reports and annual occupational radiation exposure reports.

   [[Page 51385]]

   Date of issuance: August 16, 2005.

   Effective date: As of the date of issuance and shall be implemented within 60 days.

   Amendment No.: 136.

   Facility Operating License No. NPF-58: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: May 10, 2005 (70 FR 24651).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 16, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: October 14, 2004.

   Brief description of amendment: The amendment revises a technical specification surveillance requirement to change the required frequency of the reactor building spray nozzle surveillance from once every 10 years to ``following maintenance that could result in nozzle blockage.''

   Date of issuance: August 4, 2005.

   Effective date: August 4, 2005.

   Amendment No.: 219.

   Facility Operating License No. DPR-72: Amendment revises the Technical Specifications.

   Date of initial notice in Federal Register: January 18, 2005 (70 FR 2891).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 4, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendments: February 25, 2005, as supplemented June 2, 2005.

   Brief description of amendments: The amendments modify the Technical Specifications by revising the near-end-of-life moderator temperature coefficient (MTC) surveillance requirement by placing a set of conditions on core performance, which, if met, would allow conditional exemption from the required MTC measurement.

   Date of issuance: August 8, 2005.

   Effective date: As of the date of issuance and shall be implemented within 45 days.

   Amendment Nos.: 288, 270.

   Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: March 29, 2005 (70 FR 15943).

   The supplement dated June 2, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 8, 2005.

   No significant hazards consideration comments received: No.

   Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

   Date of application for amendments: September 1, 2004, as supplemented by letter dated May 17, 2005.

   Brief description of amendments: The amendments approve the use of Generation of Thermal-Hydraulic Information Containment Version 7.1 patch 1 (GOTHIC), for licensing analyses for the Prairie Island Nuclear Generating Plants to (1) evaluate the short-term peak pressure and temperature response of the containment atmosphere to large pipe breaks in high energy piping systems--the design-basis loss-of-coolant accident (LOCA) and the design-basis main steam line break, and (2) to evaluate the long-term containment response following a design-basis LOCA.

   Date of issuance: August 12, 2005.

   Effective date: As of the date of issuance and shall be implemented within 30 days.

   Amendment Nos.: 171,161.

   Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Updated Safety Analysis Report.

   Date of initial notice in Federal Register: September 28, 2004 (69 FR 57990).

   The supplemental letter contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 12, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendments: March 5, 2004.

   Brief description of amendments: The amendments revised Technical Specifications Surveillance Requirement (SR) 3.6.4.1.3 to require that only one secondary containment access door in each access opening be verified closed. In addition, SR 3.6.4.1.3 allows entry and exit access between required secondary containment zones that have a single door.

   Date of issuance: August 16, 2005.

   Effective date: As of the date of issuance and shall be implemented within 30 days.

   Amendment Nos.: 224 and 201.

   Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: April 27, 2004 (69 FR 22882).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 16, 2005.

   No significant hazards consideration comments received: No.

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

   Date of application for amendments: September 8, 2004.

   Brief description of amendments: The amendments revised the SSES 1 and 2 Technical Specification 3.8.7, ``Distribution Systems- Operating,'' to add an action note to address the potential for deenergized Class 1E battery chargers, and correct three unrelated editorial changes.

   Date of issuance: August 17, 2005.

   Effective date: As of the date of issuance, and shall be implemented within 60 days.

   Amendment Nos.: 225 and 202.

   Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: May 24, 2005 (70 FR 29798).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 17, 2005.

   No significant hazards consideration comments received: No.

   PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

   Date of application for amendment: October 1, 2004.

   Brief description of amendment: This amendment deleted the Technical Specifications (TSs) associated with hydrogen recombiners, and hydrogen and oxygen monitors.

   Date of issuance: August 9, 2005.

   [[Page 51386]]

   Effective date: As of the date of issuance, to be implemented within 60 days.

   Amendment No.: 160.

   Facility Operating License No. NPF-57: This amendment revised the TSs.

   Date of initial notice in Federal Register: March 15, 2005 (70 FR 12749).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 9, 2005.

   No significant hazards consideration comments received: No.

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

   Date of amendment request: March 15, 2005.

   Brief description of amendments: The amendments revise the Technical Specification 3.7.10, ``Control Room Emergency Filtration/ Pressurization System (CREFS).'' The revision allows a one-time extension from 24 hours to 14 days of the allowable duration of operation with control room boundary inoperable.

   Date of issuance: August 11, 2005.

   Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance.

   Amendment Nos.: 120, 120.

   Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: May 24, 2005 (70 FR 29801).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 11, 2005.

   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 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 (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 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,

   [[Page 51387]]

   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.\1\ 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 satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

   \1\ To the extent that the applications contain attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant's counsel and discuss the need for a protective order.

   Each contention shall be given a separate numeric or alpha designation within one of the following groups:

1. Technical--primarily concerns/issues relating to technical and/ or health and safety matters discussed or referenced in the applications.

2. Environmental--primarily concerns/issues relating to matters discussed or referenced in the environmental analysis for the applications.

3. Miscellaneous--does not fall into one of the categories outlined above.

   As specified in 10 CFR 2.309, if two or more petitioners/requestors seek to co-sponsor a contention, the petitioners/requestors shall jointly designate a representative who shall have the authority to act for the petitioners/requestors with respect to that contention. If a petitioner/requestor seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/ requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/ requestor a representative who shall have the authority to act for the petitioners/requestors with respect to that contention.

   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 e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition

   for leave to intervene should also be sent to the attorney for the licensee.

   Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer or 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).

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

   Date of application for amendment: August 8, 2005, as supplemented August 11, 2005.

   Brief description of amendment: The amendment revised Technical Specification 3.3.7, ``Containment Spray System,'' specifically, increasing the maximum lake water temperature limit in specification f. from 81 [deg]F to 83 [deg]F.

   Date of issuance: August 12, 2005.

   Effective date: As of the date of its issuance and shall be implemented within 5 days.

   Amendment No.: 190.

   Facility Operating License No. DPR-63: 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 August 12, 2005.

   Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

   NRC Section Chief: Richard J. Laufer.

   Dated at Rockville, Maryland, this 22nd day of August 2005.

   [[Page 51388]]

   For the Nuclear Regulatory Commission. Ledyard B. Marsh, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

   [FR Doc. 05-16979 Filed 8-29-05; 8:45 am]

   BILLING CODE 7590-01-P