Skip to Content

Notice

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

Document Details

Information about this document as published in the Federal Register.

Published Document

This document has been published in the Federal Register. Use the PDF linked in the document sidebar for the official electronic format.

Start Preamble

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 Start Printed Page 38181the 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 June 9, 2006 to June 22, 2006. The last biweekly notice was published on June 20, 2006 (71 FR 35456).

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 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 Start Printed Page 38182when 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.

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

Date of amendment request: April 28, 2006.

Description of amendment request: The proposed amendment would revise the Seabrook Station Unit No. 1 (Seabrook) Technical Specifications (TSs) consistent with the NRC-approved Revision 9 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-359, “Increased Flexibility in MODE Restraints.”

The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments adopting TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated April 28, 2006.

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

Criterion 1—

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions [are] not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by [the] proposed LCO [limiting condition of operation] 3.0.4 are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—

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

The proposed change does not involve the physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new of different kind of accident from an accident previously evaluated.

Criterion 3—

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full compliment of equipment through the conditions for not meeting the TS Limiting Conditions for Operation (LCO). The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The new change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

Based upon the reasoning presented above it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: M. S. Ross, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: Darrell J. Roberts.

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

Date of amendment request: May 30, 2006. Start Printed Page 38183

Description of amendment request: The proposed amendment would revise the Technical Specifications, deleting from Surveillance Requirement (SR) 3.3.1.15 a note which specifies that the surveillance includes “verification of Reactor Coolant System [RCS] resistance temperature detector [RTD] bypass loop flow rate.” Approval of this proposed amendment would permit the licensee to effect a plant design change, removing the RTD bypass piping and install a replacement system using fast response thermowell-mounted RTDs located in the RCS loop piping.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee provided a no significant hazards determination analysis.

The NRC staff has reviewed the licensee's analysis and performed its own as follows:

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

No. The RTD bypass system is the passive hardware associated with RCS instrumentation with control and indication functions. The RTD bypass system was not considered a precursor to any previously analyzed accident, and was not considered a factor in the scenario leading to accident consequences. The new system replacing the RTD bypass system will perform the same control and indication functions, and similarly will not be considered a precursor to any accident, or a factor affecting accident consequences in previously analyzed accident scenarios. Therefore, replacement of the existing RTD bypass system with the new system will not increase the probability of occurrence of an accident, and will not increase consequences of an accident previously evaluated.

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

No. The replacement of the existing RTD bypass with the replacement system would not create new failure modes, and the replacement system is not an initiator of any new or different kind of accident. The proposed deletion of the note in SR 3.3.1.15 does not affect the interaction of the replacement system with any system whose failure or malfunction can initiate an accident. 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?

No. Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the models and associated assumptions used to analysis the system's performance. The replacement system will continue to perform the same temperature detection function to the same level of reliability as defined in the D.C. Cook Updated Safety Analysis Report. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

The NRC staff evaluated the licensee's analysis, and based on this evaluation, the NRC staff proposes to determine that the requested amendment does not involve a significant hazards consideration.

Attorney for licensee: James M. Petro, Jr., Esquire, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: L. Raghavan.

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 amendment request: May 26, 2006.

Description of amendment request: The licensee proposed to amend each unit's Technical Specifications in accordance with Revision 4 to Technical Specifications Task Force (TSTF) Standard TS Change Traveller, TSTF-449, “Steam Generator Tube Integrity” (see 70 FR 24126). Specifically, the following Sections will be revised per TSTF-449: Section 1.1, Definitions; Section 3.4.13, Reactor Coolant System Operational LEAKAGE; Section 5.5.7, Steam Generator (SG) Program; and Section 5.6.7, Steam Generator Tube Inspection Report. Also, a new Section 3.4.17, SG Tube Integrity, will be added. The proposed changes are necessary in order to implement the guidance for the industry initiative in Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines.

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, by referencing the NRC staff's model analysis published in 70 FR 10298 (March 2, 2005). The NRC staff's model analysis is reproduced below:

Criterion 1

—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The proposed change requires a SG Program that includes performance criteria that will provide reasonable assurance that the SG tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cooldown and all anticipated transients included in the design specification). The SG performance criteria are based on tube structural integrity, accident induced leakage, and operational LEAKAGE.

A SGTR event is one of the design-basis accidents that are analyzed as part of a plant's licensing basis. In the analysis of a SGTR event, a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in the licensing basis plus the LEAKAGE rate associated with a double-ended rupture of a single tube is assumed.

For other design-basis accidents such as MSLB [main steam line break], rod ejection, and reactor coolant pump locked rotor, the tubes are assumed to retain their structural integrity (i.e., they are assumed not to rupture). These analyses typically assume that primary to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to 1 gallon per minute as a result of accident-induced stresses. The accident-induced leakage criterion introduced by the proposed changes accounts for tubes that may leak during design-basis accidents. The accident induced leakage criterion limits this leakage to no more than the value assumed in the accident analysis.

The SG performance criteria proposed change to the TS identify the standards against which tube integrity is to be measured. Meeting the performance criteria provides reasonable assurance that the SG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design-basis accident. The performance criteria are only a part of the SG Program required by the proposed change to the TS. The program, defined by NEI 97-06, Steam Generator Program Guidelines, includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring. The proposed changes do not, therefore, significantly increase the probability of an accident previously evaluated.

The consequences of design-basis accidents are, in part, functions of the DOSE EQUIVALENT 1-131 in the primary coolant and the primary to secondary LEAKAGE rates resulting from an accident. Therefore, limits are included in the plant technical specifications for operational leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure the plant is operated within its analyzed condition. The typical analysis of the limiting design-basis accident assumes that primary to secondary leak rate after the accident is 1 gallon per minute with no more than [150] gallons per day in any one SG, and that the reactor coolant activity levels of DOSE EQUIVALENT 1-131 are at the TS values before the accident.

The proposed change does not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. The proposed approach updates the current TSs and enhances the requirements for SG inspections. The proposed change does not adversely impact any other previously evaluated design-basis accident and is an improvement over the current TSs.

Therefore, the proposed change does not affect the consequences of a SGTR accident and the probability of such an accident is reduced. In addition, the proposed changes do not affect the consequences of an MSLB, rod ejection, or a reactor coolant pump locked rotor event, or other previously evaluated accident.

Start Printed Page 38184

Criterion 2

—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated.

The proposed performance based requirements are an improvement over the requirements imposed by the current technical specifications. Implementation of the proposed SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The result of the implementation of the SG Program will be an enhancement of SG tube performance. Primary to secondary LEAKAGE that may be experienced during all plant conditions will be monitored to ensure it remains within current accident analysis assumptions.

The proposed change does not affect the design of the SGs, their method of operation, or primary or secondary coolant chemistry controls. In addition, the proposed change does not impact any other plant system or component. The change enhances SG inspection requirements.

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

Criterion 3

—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety.

The SG tubes in pressurized-water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of an SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change is expected to result in an improvement in the tube integrity by implementing the SG Program to manage SG tube inspection, assessment, repair, and plugging. The requirements established by the SG Program are consistent with those in the applicable design codes and standards and are an improvement over the requirements in the current TSs.

For the above reasons, the margin of safety is not changed and overall plant safety will be enhanced by the proposed change to the TS.

Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

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

Attorney for licensee: James M. Petro, Jr., Esquire, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: L. Raghavan.

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

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

Description of amendment request: The proposed changes would revise Technical Specification (TS) Section 5.5.6, “Inservice Testing Program,” by replacing references to Section XI 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 (10 CFR) requires that the Inservice Testing (IST) Program be updated to the latest Edition and Addenda of the ASME OM Code incorporated by reference in 10 CFR 50.55a(b) 12 months before the start of the 10-year interval. Section XI 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 next 10-year interval that began March 1, 2006. In addition, the scope of frequencies specified to be within the applicability of Surveillance Requirement (SR) 3.0.2 is expanded 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.

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

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

Response: No.

The proposed changes revise the CNS [Cooper Nuclear Station] TS for the IST Program to be consistent with the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed changes incorporate revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The proposed changes do not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events. They do not involve addition or removal of any equipment, nor any design changes to the facility.

Based on the above, NPPD [Nebraska Public Power District] concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes revise the CNS TS for the IST Program to be consistent with the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed changes incorporate revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The proposed changes do not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or a change in the methods governing normal plant operation. The proposed changes will not introduce a new accident initiator, accident precursor, or malfunction mechanism. There is no change in the types or increases in the amounts of any effluent that may be released off-site, and there is no increase in individual or cumulative occupational exposure.

Based on the above NPPD concludes that these proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed changes revise the CNS TS for the IST Program to be consistent with the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3. The proposed changes incorporate revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The safety function of the affected pumps and valves will be maintained. Based on the above, NPPD concludes that these proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: David Terao. Start Printed Page 38185

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

Date of amendment request: April 25, 2006.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) requirements for mode change limitations in TSs 3.0.4 and 4.0.4, using the CLIIP described in the Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) change, TSTF-359, Revision 9.

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

The proposed change allows entry into a MODE while relying on ACTIONS. Being in an ACTION is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on ACTIONS as allowed by the proposed LCO [limiting condition of operation] 3.0.4 are no different than the consequences of an accident while relying on ACTIONS for other reasons, such as equipment inoperability. Therefore, the consequences of an accident previously evaluated are not significantly increased by this change. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the change 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 (no new or different type of equipment will be installed). Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated; there is no change to the design basis.

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

The proposed change allows entry into a MODE or other specified conditions in the Applicability while relying on ACTIONS. The Technical Specifications allow operation of the plant without a full complement of equipment. The risk associated with this allowance is managed by the imposition of ACTIONS and Completion Times. The net effect of ACTIONS and Completion Times on the margin of safety is not considered significant. The proposed change does not change the ACTIONS or Completion Times of the Technical Specifications. The proposed change allows the ACTIONS and Completion Times to be used in new circumstances. However, this use is predicated on an assessment that focuses on managing plant risk. In addition, most current allowances to utilize the ACTIONS and Completion Times that do not require risk assessment are eliminated. As a result, the net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff 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 Branch Chief: Darrell J. Roberts.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of amendment requests: February 28, 2006.

This revised amendment request completely supercedes the licensee's submittal of December 17, 2004. Likewise, the biweekly Federal Register (FR) notice—notice of consideration of issuance of amendments to facility operating licenses, proposed no significant hazards consideration determination, and opportunity for a hearing, which was published in the FR on January 18, 2005 (70 FR 2897) is being superceded by the publication of this biweekly FR notice.

Description of amendment requests: The proposed amendment revises Technical Specifications (TSs) 3.8.1, “AC [alternating current] Sources—Operating,” 3.8.4, “DC [direct current] Sources—Operating,” 3.8.5, “DC Sources—Shutdown,” 3.8.6, “Battery Cell Parameters,” 3.8.7, “Inverters—Operating,” and 3.8.9, “Distribution Systems—Operating.” This change will also add a new Battery Monitoring and Maintenance Program, Section 5.5.2.16.

The proposed TS changes will provide operational flexibility supported by DC electrical subsystem design upgrades that are in progress. These upgrades will provide increased capacity batteries, additional battery chargers, and the means to cross-connect DC subsystems while meeting all design battery loading requirements. With these modifications in place, it will be feasible to perform routine surveillances as well as battery replacements online.

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. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to Technical Specifications (TS) 3.8.4 and 3.8.6 would allow extension of the Completion Time (CT) for inoperable Direct Current (DC) distribution subsystems to manually cross-connect DC distribution buses of the same safety train of the operating unit for a period of 30 days. Currently the CT only allows for 2 hours to ascertain the source of the problem before a controlled shutdown is initiated. Loss of a DC subsystem is not an initiator of an event. However, complete loss of a Train A (subsystems A and C) or Train B (subsystems B and D) DC system would initiate a plant transient/plant trip.

Operation of a DC Train in cross-connected configuration does not affect the quality of DC control and motive power to any system. Therefore, allowing the cross-connect of DC distribution systems does not significantly increase the probability of an accident previously evaluated in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR).

The above conclusion is supported by Probabilistic Risk Assessment (PRA) evaluation which encompasses all accidents, including UFSAR Chapter 15.

Modification to the Frequency for Surveillance Requirements in TS 3.8.4, 3.8.5, and 3.8.6 are consistent with previously described recommendations. Enhancements from TSTF-360, Rev. 1 and IEEE 450-2002 have been incorporated into Limiting Conditions for Operation (LCOs) 3.8.4, 3.8.5, and 3.8.6. These changes do not impact the probability or consequences of an accident previously evaluated.

Further changes are made of an editorial nature or provide clarification only. For example, discussions regarding electrical ‘Trains' and ‘Subsystems' will be in more conventional terminology. LCOs affected by editorial changes include 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9.

The changes being proposed in the TS do not affect assumptions contained in other safety analyses or the physical design of the plant, nor do they affect other Technical Specifications that preserve safety analysis assumptions.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously analyzed.

2. Will operation of the facility in accordance with this proposed change create the possibility of new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change modifies surveillances and LCOs for batteries and chargers to meet the requirements of IEEE Start Printed Page 38186450-2002 whose intent is to maintain the same equipment capability as previously assumed in our commitment to IEEE 450-1980.

The proposed change will allow the cross-tie of DC subsystems and allow extension of the CT for an inoperable subsystem to 30 days. Failure of the crosstied DC buses and/or associated battery(ies) is bounded by existing evaluations for the failure of an entire electrical train.

Swing battery chargers are added to increase the overall DC system reliability. Administrative and mechanical controls will be in place to ensure the design and operation of the DC systems continue to meet the UFSAR design basis.

LCOs 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9 revisions are editorial clarifications and do not affect plant design.

Therefore, operation of the facility in accordance with this proposed change will not create the possibility of new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No.

Changes in accordance with IEEE 450-2002 and TSTF-360, Rev. 1 maintain the same level of equipment performance stated in the UFSAR and the current Technical Specifications.

Swing battery chargers are added to increase the overall DC system reliability. Administrative and mechanical controls will be in place to ensure the design and operation of the DC systems continue to meet the UFSAR design basis.

The addition of the DC cross-tie capability proposed for LCO 3.8.4 has been evaluated, as described previously, using PRA and determined to be of acceptable risk as long as the duration while cross-tied is limited to 30 days. An LCO has been included as part of this proposed change to ensure that plant operation, with DC buses cross-tied, will not exceed 30 days.

All remaining changes are editorial.

Therefore, operation of the facility in accordance with the proposed amendment would 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 requests involve no significant hazards consideration.

Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: David Terao.

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.

Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: March 1, 2006, supplemented April 26, 2006.

Brief description of amendments: The amendments revised the Technical Specifications to reconcile the criticality requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, and 10 CFR part 72 for loading and unloading dry spent fuel pool canisters in the spent fuel pool.

Date of Issuance: June 15, 2006.

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

Amendment Nos.: 351/353/352.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the Licenses and Technical Specifications.

Date of initial notice in Federal Register: April 11, 2006 (71 FR 18373).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 15, 2006.

The supplement dated April 26, 2006, provided clarifying information that did not change the scope of the original application and the initial proposed no significant hazards consideration determination.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

Date of application for amendment: April 22, 2005.

Brief description of amendment: The amendments revise the surveillance requirements (SRs) for Technical Specification 3.3.5, “Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.” Specifically, a note was added to IP2 SR 3.3.5.2 to indicate that the verification of the setpoint is not required for the 480 volt (V) bus degraded voltage function when performing the trip actuating device operational test. A similar note was added to IP3 SR 3.3.5.1 for the 480 V degraded voltage and undervoltage functions.

Date of issuance: June 7, 2006.

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

Amendment No.: 247 and 231.

Facility Operating License Nos. DPR-26 and DPR-64: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 7, 2005 (70 FR 33213).

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

No significant hazards consideration comments received: No. Start Printed Page 38187

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

Date of application for amendments: April 30, 2004, as supplemented by letters dated December 17, 2004; June 30, 2004; July 5, 2005; September 30, 2005; and June 1, 2006.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.3.1.3, “Oscillation Power Range Monitor (OPRM) Instrumentation”; TS 3.4.1, “Recirculation Loops Operating”; and TS 5.6.5, “Core Operating Limits Report (COLR)”; to insert a new TS section for the ORPM instrumentation, delete the current thermal-hydraulic instability administrative requirements, and add the appropriate references for the OPRM trip setpoints and methodology.

Date of issuance: June 13, 2006.

Effective date: As of the date of issuance and shall be implemented within 150 days.

Amendment Nos.: 177/163.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Technical Specifications and License.

Date of initial notice in Federal Register: June 8, 2004 (69 FR 32073).

The December 17, 2004; June 30, 2004; July 5, 2005; September 30, 2005; and June 1, 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 June 13, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 25, 2005.

Brief description of amendments: The amendments deleted the sections of the Facility Operating Licenses that require reporting of violations of the requirements in Sections 2.C and 2.E of the Facility Operating Licenses.

Date of issuance: June 14, 2006.

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

Amendment Nos.: 178/164.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the License.

Date of initial notice in Federal Register: April 26, 2005 (70 FR 21456).

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

No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-335 and 50-389, St. Lucie Plants, Units 1 and 2, St. Lucie County, Florida

Date of application for amendments: April 21, 2005.

Brief description of amendments: The amendments revised Technical Specifications (TSs) to adopt seven TS Task Force (TSTF) generic changes (TSTF nos. 5, 65, 101, 258, 299, 308, and 361) that delete redundant safety limit violation notification requirements; adopt use of generic titles for utility positions; change the auxiliary feedwater pump test requirements to be consistent with the inservice test program; remove redundant requirements and add other requirements to Section 5.0, Administrative Controls; clarify the meaning of “refueling cycle” for system integrated leak test intervals in the Primary Coolant Sources Outside Containment program; clarify the requirements regarding the frequency of testing for cumulative and projected dose contributions from radioactive effluents; and add a note to the residual heat removal (RHR) requirements during Mode 6 low water level operations that allows one required RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is operable and in operation.

Date of issuance: June 19, 2006.

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

Amendment Nos: 199 and 146.

Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the TSs.

Date of initial notice in Federal Register: July 5, 2005 (70 FR 38720).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 19, 2006.

No significant hazards consideration comments received: No.

Nuclear Management Company (NMC), LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of application for amendments: November 12, 2004, as supplemented by letters dated January 30 and March 6, 2006.

Brief description of amendments: The amendments revise Technical Specification 5.5.7, “Inservice Testing Program” to update the references to the American Society of Mechanical Engineers Code and certain associated periodicities for inservice testing activities, consistent with the requirements of 10 CFR 50.55a.

Date of issuance: June 8, 2006.

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

Amendment Nos.: 222 and 228.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: January 17, 2006 (71 FR 2592).

The January 30, 2006, supplement withdrew a portion of the original request and the March 6, 2006, supplement contained clarifying information.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 8, 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 Start Printed Page 38188of 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.[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.

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.Start Printed Page 38189

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).

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

Date of amendment request: June 7, 2006, as supplemented by letters dated June 8, and June 9, 2006.

Description of amendment request: The amendment revised Technical Specification (TS) 3.6.5.1, “Containment Enclosure Emergency Air Cleanup Systems,” to increase the TS allowed outage time with one inoperable enclosure air handling fan EAH-FN-31B from 7 days to 14 days, on a one-time basis.

Date of issuance: June 9, 2006.

Effective date: As of its date of issuance and shall be implemented prior to the expiration of the current 7-day allowed outage time entered on June 4, 2006, for fan EAH-FN-31B.

Amendment No.: 111.

Facility Operating License No. NPF-86: The amendment revised the TSs.

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 June 9, 2006.

Attorney for licensee: M. S. Ross, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: Darrell J. Roberts.

Start Signature

Dated at Rockville, Maryland, this day of June 26, 2006.

For the Nuclear Regulatory Commission.

Catherine Haney,

Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

End Signature End Preamble

Footnotes

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.

Back to Citation

[FR Doc. 06-5899 Filed 7-3-06; 8:45 am]

BILLING CODE 7590-01-P