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 September 1, 2006, to September 14, 2006. The last biweekly notice was published on September 12, 2006 (71 FR 53715).

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, Rulemaking, Directives and Editing 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 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 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona

Date of amendments request: July 20, 2006.

Description of amendments request: The proposed amendments would revise Technical Specification 3.1.6, “Shutdown Control Element Assembly (CEA) Insertion 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.

Safety analyses require that the shutdown CEAs insert into the core at least 90% within 4 seconds of the safety signal initiating the shutdown sequence with the assumption that the shutdown CEAs' starting positions are at 150 inches withdrawn. This assumption will not be altered with the new proposed withdrawal limit.

The positioning of control rods (shutdown CEAs) to a new limit of ≥147.75 inches withdrawn is not a precursor to any accident analyzed at Palo Verde nor do these conditions affect any accident precursor; thus, initial control rod position does not change the probability of an accident previously evaluated.

To assess the effect control rod position would have on the safety analyses with the rods positioned at the new limit, several events and specific parameters were analyzed. The events were chosen because of their sensitivity to rod position. The specific parameters were analyzed to determine if, with the rods positioned at the new limit, the power distribution in the core was still within the assumptions made in the safety analyses.

Since none of the related safety analyses resulted in a significant change in the previously calculated values and the limiting parameters associated for those analyses were not exceeded, the consequences of these accidents remain unchanged. Therefore, the new insertion limit for the shutdown CEAs will not increase the consequences of any accident analyzed in our licensing bases documents.

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 accident from any accident previously evaluated?

Response: No.

PVNGS [Palo Verde Nuclear Generating Station] licensing bases documents describe the design function of the control rods as components that include a positive means (gravity) for inserting the control rods and are capable of reliably controlling the nuclear reactor to assure that under conditions of normal operation, including anticipated accidents, fuel design limits are not exceeded. The proposed amendment, new control rod (shutdown CEA) insertion limit, does not create the possibility of a new or different kind of accident from any accident previously evaluated nor does it affect the control rods ability to perform its design function.

Control rods placed at the new insertion limit will not cause fuel design limits to be exceeded during normal operations or accidents. Placing the control rods at the new insertion limit in no way impedes their insertion due to gravity. These CEAs are tested to ensure that they will insert greater than 90% into the core in less than 4 seconds from a completely withdrawn position (150 inches) and this requirement will continue to be met.

Establishing a new insertion limit for the control rods does not modify any of the existing components or systems used to position the control rods. The new insertion limit will also satisfy the assumptions made in the safety analyses.

In conclusion, the new insertion limit stills [sic] allows the control rods to fulfill their design function and does not create a new or different accident than is already described in the licensing bases documents. 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 amendment, new shutdown CEA insertion limit, does not involve a reduction in the margin of safety. The new shutdown CEA insertion limit does not affect any of the limits used to determine the acceptability of newly designed cores. The safety analyses in the licensing bases documents remain acceptable when this new (more restrictive) shutdown CEA insertion limit is applied. Additionally, the design basis of the control rods is unaffected by the new insertion limit. The design function of the control rods is to provide a positive means (gravity) for inserting the control rods and is capable of reliably controlling the nuclear reactor to assure that under conditions of normal operation, including anticipated accidents, fuel design limits are not exceeded. Since the bounding safety analyses limits used remain the same and the control rod design basis is unaffected, the fuel design limits associated with the clad material; which houses the fuel; and the design limits of the coolant system; which houses the fuel assemblies; remain unchanged. Therefore, the margin of safety is not reduced.

In conclusion, since the bounding limits used for safety analyses are unaffected by the new shutdown CEA insertion limit, the safety limits associated with the fuel and the coolant system remain unchanged. The design basis on the control rods is to ensure the fuel safety limits are not exceeded and since they remain unchanged, the design basis is still achieved. Therefore, there is no reduction in the margin of safety.

Therefore, APS [Arizona Public Service] has concluded that the proposed license amendment request does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Janet S. Mueller, Director, Law Department, Arizona Public Service Company, P.O. Box 52034, Mail Station 8695, Phoenix, Arizona 85072-2034.

NRC Branch Chief: David Terao.

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

Date of amendment request: August 2, 2006.

Description of amendment request: The proposed change would delete Waterford 3 Technical Specification Surveillance Requirement (SR) 4.6.1.7.2. This SR is the augmented testing requirement for containment purge supply and exhaust isolation valves with resilient seal materials and allows the surveillance intervals to be set in accordance with the Containment Leakage Rate Testing Program.

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

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

Response: No.

This change deletes the augmented testing requirement for these containment isolation valves and allows the surveillance intervals to be set in accordance with the Containment Leakage Rate Testing Program. This change does not affect the system function or design. The purge valves are not an initiator of any previously analyzed accident. Leakage rates do not affect the probability of the occurrence of any accident. Operating history has demonstrated that the valves do not degrade and cause leakage as previously anticipated. Because these valves have been demonstrated to be reliable, these valves can be expected to perform the containment isolation function as assumed in the accident analyses. Therefore, there is no significant increase in the consequences of any previously evaluated accident.

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

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

Response: No.

Extending the test intervals has no influence on, nor does it contribute in any way to, the possibility of a new or different kind of accident or malfunction from those previously analyzed. No change has been made to the design, function or method of performing leakage testing. Leakage acceptance criteria have not changed. No new accident modes are created by extending the testing intervals. No safety-related equipment or safety functions are altered as a result of this change.

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 only margin of safety that has the potential of being impacted by the proposed change involves the offsite dose consequences of postulated accidents which are directly related to the containment leakage rate. The proposed change does not alter the method of performing the tests nor does it change the leakage acceptance criteria. Sufficient data has been collected to demonstrate these resilient seals do not degrade at an accelerated rate.

Because of this demonstrated reliability, this change will provide sufficient surveillance to determine an increase in the unfiltered leakage prior to the leakage exceeding that assumed in the accident analysis.

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: N.S. Reynolds, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817.

NRC Branch Chief: David Terao.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of amendment request: May 11, 2006.

Description of amendment request: The proposed change would revise Technical Specification (TS) 3.1.7, “Standby Liquid Control (SLC) System,” to change the minimum required SLC pump discharge pressure specified in surveillance requirement (SR) SR 3.1.7.7 from 1235 psig to 1320 psig. This change is in response to Nuclear Regulatory Commission Information Notice 2001-13, “Inadequate Standby Liquid Control System Relief Valve Margin.”

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

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

Response: No.

The proposed change revises the surveillance requirements for the SLC system to correspond to the maximum expected pressure in the reactor pressure vessel for an ATWS [anticipated transient without scram] event. This proposed increase in the specified SLC pump discharge pressure involves only the SLC system. No other NMP2 structures, systems, or components are affected. The SLC system is provided to mitigate ATWS events and, as such, is not considered to be an initiator of an ATWS event or any other analyzed accident. The revised TS surveillance requirement, and the associated change to the SLC pump discharge relief valve set pressure (not described in the TS), neither reduce the ability of the SLC system to respond to and mitigate an ATWS event nor increase the likelihood of a system malfunction that could increase the consequences of an accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to the SLC pump TS surveillance requirement, and the associated change to the SLC pump discharge relief valve set pressure (not described in the TS), are consistent with the functional requirements of the ATWS rule (10 CFR 50.62). The proposed change does not involve the installation of any new or different type of equipment, does not introduce any new modes of plant operation, and does not change any methods governing normal plant operation. The proposed change does not introduce any new accident initiators, and therefore 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 change does not alter any assumptions, initial conditions or results from any accident analyses. The proposed change to the SLC pump TS surveillance requirement, and the associated changes to the SLC pump discharge relief valve set pressure (not described in the TS), are consistent with the functional requirements of the ATWS rule (10 CFR 50.62). The ability of the SLC system to respond to and mitigate an ATWS event is not affected. 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: Mark J. Wetterhahn, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Richard J. Laufer.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of amendment request: August 11, 2006.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.3.2.1, “Control Rod Block Instrumentation,” to revise the number of startups allowed with the rod worth minimizer (RWM) inoperable from one per calendar year to two per operating cycle (approximately 2 years).

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

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

Response: No.

The proposed change redefines the frequency at which plant startup is permitted without using the RWM. The relevant design basis accident is the control rod drop accident (CRDA), which involves multiple failures to initiate the event. This administrative change does not increase the probability of occurrence of any of the failures that are necessary for a CRDA to occur. Use of the RWM or the alternate use of a qualified human checker to ensure the correct control rod withdrawal sequence is not in itself an accident initiator, and redefining the startup allowance frequency does not involve any plant hardware changes or new operator actions that could serve to initiate a CRDA. The proposed change will have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. Also, since the banked position withdrawal sequence (BPWS) will continue to be enforced by either the RWM or verification by a second qualified individual, the initial conditions of the CRDA radiological consequence analysis presented in the U[F]SAR [Updated Final Safety Analysis Report] are not affected. Therefore, there will be no 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 and will not result in a change to the design function or operation of any structure, system, or component that is used for accident mitigation. The proposed redefinition of the frequency at which plant startup is permitted without using the RWM does not result in any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing basis. This administrative change does not affect the ability of safety-related systems and components to perform their intended safety functions. Therefore, the proposed change 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 a margin of safety?

Response: No.

The proposed change redefines the frequency at which plant startup is permitted without using the RWM. This administrative change does not affect the overall frequency of use of the allowance. The proposed change will have no adverse affect on plant operation or equipment important to safety. The relevant design basis accident is the control rod drop accident (CRDA), which involves multiple failures to initiate the event. The CRDA analysis consequences and related initial conditions remain unchanged when invoking the proposed change. The plant response to the CRDA will not be affected and the accident mitigation equipment will continue to function as assumed in the accident analysis. Therefore, there will be no significant reduction in a margin of safety.

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

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Richard J. Laufer.

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

Date of amendment request: July 5, 2006.

Description of amendment request: The proposed change would revise the main control room (MCR) and emergency switchgear room (ESGR) air conditioning system (ACS) Technical Specifications (TSs) to reflect the completion of permanent modifications to the equipment and associated power supply configuration. The revisions include the addition of requirements and/or action statements addressing the inoperability of two or more air handling units (AHUs) on a unit, as well as AHUs powered from an H emergency bus. The proposed change, paralleling requirements in the Improved Technical Specifications (ITS), also adds MCR and ESGR ACS requirements during refueling operations and irradiated fuel movement in the fuel building. In addition, the proposed change clarifies the service water (SW) requirements for the ACS chillers that serve the MCR and ESGRs.

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

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

The proposed change does not impact the condition or performance of any plant structure, system, or component. The proposed change does not affect the initiators of analyzed events or the assumed mitigation of accident or transient events. No physical changes to the ACS or SW System are involved, and accident operation of the ACS will not change. As a result, the proposed change to the Surry Technical Specifications does not involve any significant increase in the probability or the consequences of any accident or malfunction of equipment important to safety previously evaluated since neither accident probabilities nor consequences are being affected by this proposed change.

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

The proposed change does not involve a physical alteration of the plant or a change in the methods used to respond to plant transients. No new or different equipment is being installed, and no installed equipment is being removed. There is no alteration to the parameters with which the plant is normally operated or in the setpoints, which initiate protective or mitigative actions. The ACS will continue to perform its required function. Consequently, no new failure modes are introduced by the proposed change. Therefore, the proposed change to the Surry Technical Specifications does not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any previously evaluated.

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

The proposed TS change does not impact any plant structure, system, or component that is relied upon for accident mitigation. Margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. Since ACS performance is not affected by the proposed change, the ACS will continue to be available to perform its required function. Furthermore, the change does not affect the condition or performance of structures, systems, or components relied upon the accident mitigation or any safety analysis assumptions. Therefore, the proposed change to the Surry Technical Specifications 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: Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Branch Chief: Evangelos C. Marinos.

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

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: July 20, 2006.

Brief description of amendment request: The proposed amendment would revise the Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Technical Specifications (TS) 5.5.9, “Steam Generator (SG) Tube Surveillance Program,” to incorporate changes in the SG inspection scope for VEGP, Unit 1 during Refueling Outage 13 and the subsequent operating cycle, and VEGP Unit 2 during Refueling Outage 12 and the subsequent operating cycle. The proposed changes modify the inspection requirements for portions of SG tubes within the tubesheet region of the SGs.

Date of publication of individual notice in Federal Register: July 31, 2006 (71 FR 43225).

Expiration date of individual notice: 30-day August 30, 2006; 60-day, September 29, 2006.

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 (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, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: October 18, 2005, as supplemented by letter dated May 26, 2006.

Brief description of amendment: The amendment revised the Oyster Creek Nuclear Generating Station Technical Specifications (TSs) Surveillance Requirement (SR) 4.4.B.1 to provide an alternative means for testing the electromatic relief valves located on the main steam system. The revised SR allows demonstration of the capability of the valves to perform their function without requiring that the valves be cycled with steam pressure while installed.

Date of Issuance: September 1, 2006.

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

Amendment No.: 260.

Facility Operating License No. DPR-16: The amendment revised the TSs.

Date of initial notice in Federal Register: December 20, 2005 (70 FR 75490). The May 26, 2006, 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.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated September 1, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 27, 2004.

Brief description of amendments: The amendments revised the facility operating licenses by removal of Section 2.E, that lists reporting requirements with regard to Maximum Power Level, Updated, Fire Protection, Protection of the Environment (Unit 2 only) and Physical Protection.

Date of issuance: September 7, 2006.

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

Amendment Nos.: 233 and 215.

Renewed Facility Operating License Nos. NPF 9 and NPF-17: Amendments revised the licenses.

Date of initial notice in Federal Register: July 5, 2005 (70 FR 38717).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 7, 2006.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: April 27, 2005, as supplemented by letters dated November 22, 2005, and August 1, 2006. The August 1, 2006, submittal reduced the scope of the changes to only revise Technical Specification Limiting Condition for Operation 3.8.4, “DC Sources-Operating.”

Brief description of amendment: The amendment revises the Technical Specifications to allow a battery charger to be out of service for up to 7 days.

Date of issuance: September 14, 2006.

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

Amendment No.: 286.

Facility Operating License No. DPR-59: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: July 19, 2005 (70 FR 41444). The November 22, 2005, and August 1, 2006, supplements 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 amendment is contained in a Safety Evaluation dated September 14, 2006.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2

Will County, Illinois

Date of application for amendment: February 15, 2005, as supplemented by letters dated November 28 and December 9, 2005 (two letters), and January 27, February 13, March 17 and July 14, 2006.

Brief description of amendment: The amendments fully implement an alternative source term.

Date of issuance: September 8, 2006.

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

Amendment Nos.: 147, 147, 140 and 140.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revised the Technical Specifications and License.

Date of initial notice in Federal Register: May 10, 2005 (70 FR 24650). The November 28 and December 9, 2005 (two letters), and January 27, February 13, March 17 and July 14, 2006 supplements, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 8, 2006.

No significant hazards consideration comments received: No.

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

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

Date of application for amendments: October 10, 2002, as supplemented by letters dated March 21, March 28, August 4, September 15 and October 31, 2003, and June 30, August 6, September 3, September 10, September 22, November 2 and November 5, 2004, and March 3, August 22, September 3 and September 27, 2005, and February 17 and May 25, 2006.

Brief description of amendments: The amendments adopt the alternative source term methodology as prescribed in Title 10 to the Code of Federal Regulations Section 50.67.

Date of issuance: September 11, 2006.

Effective date: As of the date of issuance and shall be implemented within 180 days.

Amendment Nos.: 221/212, 233/229.

Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. The amendments revised the Technical Specifications, Surveillance Requirements and Licenses.

Date of initial notice in Federal Register: August 19, 2003 (68 FR 49816). The supplements dated March 21, March 28, August 4, September 15 and October 31, 2003, and June 30, August 6, September 3, September 10, September 22, November 2, and November 5, 2004, and March 3, August 22, September 3 and September 27, 2005, and February 17 and May 25, 2006, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 11, 2006.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

Date of application for amendment: April 10, 2006, as supplemented by letters dated April 12, 13 (2 letters), and June 27, 2006.

Brief description of amendment: The amendment revised Surveillance Requirement 3.8.1.11 of the DCCNP-1 Technical Specifications, raising the diesel generator load rejection voltage test limit from 5000 volts to 5350 volts.

Date of issuance: September 1, 2006.

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

Amendment No.: 295.

Facility Operating License No. DPR-58: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 1, 2006 (71 FR 43534). The supplemental letters 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 amendment is contained in a Safety Evaluation dated September 1, 2006.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 29, 2005, as supplemented by letters dated January 16, and April 7, 2006.

Brief description of amendment: The amendment eliminated operability requirements for secondary containment, secondary containment isolation valves, the standby gas treatment system, and secondary containment isolation instrumentation when handling irradiated fuel that has decayed for 24 hours since critical reactor operations, and when performing core alterations. Similar technical specification relaxations are granted for the Control Room Emergency Filter System and its initiation instrumentation after a decay period of 7 days.

Date of issuance: September 5, 2006.

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

Amendment No.: 222.

Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 3, 2006 (71 FR 149). The supplements dated January 16 and April 17, 2006, 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 September 5, 2006.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 7, 2006, as supplemented by letter dated May 10, 2006.

Brief description of amendment: The amendment revised Technical Specification (TS) Section 5.5.6, “Inservice Testing Program,” by replacing references to Section Xl of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code with ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). Section 50.55a of Title 10 of the Code of Federal Regulations (CFR) requires that the Inservice Testing (IST) Program be updated to the latest Edition and Addenda of the Code incorporated by reference in 10 CFR 50.55a(b) 12 months before the start of the applicable 10-year interval. Section Xl of the ASME Boiler and Pressure Vessel Code has been replaced with the ASME OM Code as the code of reference for IST programs. Thus, the ASME OM Code is the code of reference for the IST Program for the 10-year interval that began March 1, 2006. In addition, the amendment expanded the scope of frequencies specified to be within the applicability of Surveillance Requirement (SR) 3.0.2 by adding mention of other normal and accelerated frequencies specified in the IST Program. This will eliminate any confusion regarding the applicability of SR 3.0.2 to IST Program Frequencies.

Date of issuance: September 6, 2006.

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

Amendment No.: 223.

Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register : July 5, 2006 (71 FR 38184).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 6, 2006.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 1, 2005, as supplemented on September 16, 2005, November 15, 2005, December 14, 2005, February 16, 2006, and July 6, 2006.

Brief description of amendment: The amendment revises the Updated Safety Analysis Report, Section 14.10, “Malfunctions of the Feedwater System,” to describe an existing Emergency Operating Procedure operator action to isolate the steam generator blowdown within 15 minutes of a reactor trip during a loss-of-main feedwater event.

Date of issuance: September 11, 2006.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment No.: 242.

Renewed Facility Operating License No. DPR-40: The amendment revised the Updated Safety Analysis Report.

Date of initial notice in Federal Register : August 2, 2005 (70 FR 44403). The September 16, 2005, November 15, 2005, December 14, 2005, February 16, 2006, and July 6, 2006, supplemental letters provided information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a safety evaluation dated September 11, 2006.

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: September 26, 2005, as supplemented by letter dated June 28, 2006.

Brief description of amendments: The proposed amendments revised the Salem Technical Specifications (TSs) to eliminate certain Surveillance Requirements (SRs) for containment isolation valves. The changes deleted SR 4.6.3.1.1 and SR 4.6.3.1 for Salem Unit Nos. 1 and 2, respectively. These SRs require a complete valve stroke and stroke time measurement when a valve is returned to service after maintenance, repair, or replacement work. The changes are intended to minimize unnecessary testing and plant transients. Other Salem TS containment isolation valve SRs ensure that the valves remain operable.

Date of issuance: August 31, 2006.

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

Amendment Nos.: 274 and 255.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register : July 18, 2006 (71 FR 40739).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 31, 2006.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of application for amendments: January 27, 2005, as supplemented by letters dated September 30, 2005, and January 25 and May 5, 2006.

Brief description of amendments: The amendments revised the Technical Specifications by extending the surveillance test interval for components of the reactor protection system.

Date of issuance: September 1, 2006.

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

Amendment Nos.: 145 and 125

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register : November 8, 2005 (70 FR 67751). The supplements dated September 30, 2005, and January 25 and May 5, 2006, provided clarifying information that did not change the scope of the January 27, 2005, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 1, 2006.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of application for amendments: July 20, 2006, as supplemented by letter dated August 4, 2006.

Brief description of amendments: The amendments revised Technical Specification 5.5.9, “Steam Generator (SG) Tube Surveillance Program,” regarding the required SG inspection scope for Vogtle, Unit 1, during Refueling Outage 13 and the subsequent operating cycle and Vogtle, Unit 2, during Refueling Outage 12, and the subsequent operating cycle. The proposed changes modify the inspection requirements for portions of the SG tubes within the hot leg tubesheet region of the SGs.

Date of issuance: September 12, 2006.

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

Amendment Nos.: 146 and 126.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register : July 31, 2006 (71 FR 43225). The supplement dated August 4, 2006, provided clarifying information that did not expand the scope of the July 20, 2006, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 12, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 10, 2006 as supplemented by letters April 14, August 1, September 5 and 14, 2006.

Description of amendment request: The amendments revised Technical Specifications 3.3.1.1 and 3.3.5.1 to specify the methodology used for determining, setting, and evaluating as-found setpoints for drift-susceptible instruments that are necessary to ensure compliance with a Safety Limit or are critical in ensuring the fuel peak cladding temperature acceptance criterion are met.

Date of issuance: September 14, 2006.

Effective date: Date of issuance, to be implemented within 90 days.

Amendment Nos.: 257, 296 and 254.

Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register : March 28, 2006 (71 FR 15487). The supplements dated April 14, August 1, September 5 and 14, 2006, 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 determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 14, 2006.

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 amendments request: December 16, 2005, as supplemented by letter dated June 7, 2006.

Brief description of amendments: The amendments revised the steam generator tube surveillance program technical specifications (TSs) to be consistent with TS Task Force (TSTF) traveler TSTF-449, Revision 4, “Steam Generator Tube Integrity.”

Date of issuance: September 12, 2006.

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

Amendment Nos.: 128/128.

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

Date of initial notice in Federal Register : March 14, 2006 (71 FR 13181). The supplement dated June 7, 2006, 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 amendments is contained in a Safety Evaluation dated September 12, 2006.

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

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).

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

Date of amendment request: June 2, 2006.

Description of amendment request: The amendment deleted Technical Specifications (TSs) 4.3.1.2b and TS 4.3.1.2c of the FCS TSs. The amendment also made an administrative change to TS 4.3.1.2 to correct the current wording of TS 4.3.1.2 and TS 4.3.1.2d. TS 4.3.1.2 implied that more than one new fuel storage rack at FCS is installed when there is actually only one new fuel storage rack. In addition, Omaha Public Power District (OPPD) will complete additional procedural enhancements of administrative controls for compliance with 10 CFR 50.68(b)(2) and (b)(3) prior to receipt of new fuel for the 2006 Refueling.

Date of issuance: June 27, 2006.

Effective date: The license amendment is effective as of its date of issuance and shall be implemented within 7 days of issuance. OPPD will complete additional enhancements of administrative controls for compliance with 10 CFR 50.68(b)(2) and (b)(3) prior to receipt of new fuel for the 2006 Refueling.

Amendment No.: 240.

Renewed Facility Operating License No. DPR-40: Amendment revised the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC):

Yes. Omaha World-Herald on June 11, 2006. 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 August 31, 2006.

Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: David Terao.