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 Start Printed Page 26996determination 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 April 14, 2006 to April 27, 2006. The last biweekly notice was published on April 25, 2006 (71 FR 23952).

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 Start Printed Page 26997determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: January 26, 2006.

Description of amendment request: The proposed amendment would update the list of Nuclear Regulatory Commission-approved documents specified in the Technical Specifications that describe the analytical methods used to determine the core operating limits.

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

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

Response: No.

The proposed amendment adds a new document (No. 16) to TS 6.9.1.8 b to complement the list of documents used to determine the core operating limits. These documents have been previously reviewed and approved by the NRC. It also changes the word “minimum” to “maximum” in TS 5.3.1 to correctly state the limit on nominal average enrichment of reload fuel. This change restores TS 5.3.1 wording to the wording previously approved by the NRC in Amendment 274. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes have no effect on the consequence of any analyzed accident since the changes do not affect the function of any equipment credited for accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes do not modify any plant equipment and there is no impact on the capability of existing equipment to perform its intended functions. No system setpoints are being modified and no changes are being made to the method in which plant operations are conducted. No new failure modes are introduced by the proposed change. The proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed amendment adds a new document (No. 16) to TS 6.9.1.8 b to complement the list of documents used to determine the core operating limits. These documents have been previously reviewed and approved by the NRC. It also changes the word “minimum” to “maximum” in TS 5.3.1 to correctly state the limit on nominal average enrichment of reload fuel. This change restores TS 5.3.1 wording to the wording previously approved by the NRC in Amendment 274. The proposed changes have no impact on plant equipment operation. The proposed changes do not revise any setpoints nor do they change the acceptance criteria used in the accident analyses. Therefore, the proposed changes will not result in a 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, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Branch Chief: Darrell J. Roberts.

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

Date of amendment request: March 28, 2006.

Description of amendment request: The proposed amendment would delete the license condition, Section 2.F of Facility Operating License No. NPF-49, which requires reporting of violations of the requirements in Section 2.C of Facility Operating License No. NPF-49. The change is consistent with the notice published in the Federal Register on November 4, 2005, as part of the consolidated line item improvement process.

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:

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

Response: No. Start Printed Page 26998

The proposed change involves the deletion of a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not significantly increase 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?

Response: No.

The proposed change is administrative in that it deletes a reporting requirement. The change does not add new plant equipment, change existing plant equipment, or affect the operating practices of the facility. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change deletes a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not involve a significant reduction in a margin of safety.

Based on the above, the NRC staff proposes that the change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

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

NRC Branch Chief: Darrell J. Roberts.

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

Date of amendment request: June 15, 2005.

Description of amendment request: The proposed amendments would revise the Technical Specifications to eliminate the out of date requirements associated with the completion of the Keowee Refurbishment modifications on both Keowee Hydro Units.

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. Involve a significant increase in the probability or consequences of an accident previously evaluated:

The proposed change to the Oconee Technical Specification (TS) 3.8.1 removes out of date requirements associated with temporary extensions to Required Action (RA) Completion Times (CTs) that are no longer applicable because of the completion of the Keowee Refurbishment modifications on both KHUs. The proposed change also removes a Facility Operating License (FOL) License Condition that is no longer needed since the associated TS change is no longer applicable. As such, the proposed change is administrative. No actual plant equipment, operating practices, or accident analyses are affected by this change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any kind of accident previously evaluated:

The proposed change to the Oconee TSs and FOLs removes requirements associated with a temporary extension of TS 3.8.1 RA CTs that are no longer applicable because of the completion of the Keowee Refurbishment modifications on both KHUs. As such, the proposed changes are administrative. No actual plant equipment, operating practices, or accident analyses are affected by this change. No new accident causal mechanisms are created as a result of this change. The proposed change does not impact any plant systems that are accident initiators; neither does it adversely impact any accident mitigating systems. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed change does not adversely affect any plant safety limits, set points, or design parameters. The change also does not adversely affect the fuel, fuel cladding, Reactor Coolant System, or containment integrity. The proposed change eliminates requirements that are no longer applicable and is administrative in nature. Therefore, the proposed change does not involve a 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: Ms. Lisa F. Vaughn, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

NRC Branch Chief: Evangelos C. Marinos.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of amendment request: April 17, 2006.

Description of amendment request: The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed consistent with the program in place for complying with the requirements of paragraph 50.65(a)(4) of Title 10 of the Code of Federal Regulations (10 CFR). Limiting Condition for Operation (LCO) 3.0.8 is added to the TS to provide this allowance and define the requirements and limitations for its use.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff issued a notice of opportunity for comment in the Federal Register on November 24, 2004 (69 FR 68412), on possible amendments concerning TSTF-372, 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 May 4, 2005 (70 FR 23252). The licensee affirmed the applicability of the following NSHC determination in its application dated April 17, 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 a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and Start Printed Page 26999managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG [Regulatory Guide] 1.177. A bounding risk assessment was performed to justify the proposed TS changes. [The proposed LCO 3.0.8 defines limitations on the use of the provision and includes a requirement for the licensee to assess and manage the risk associated with operation with an inoperable snubber.] The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: David Terao.

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

Date of amendment request: March 20, 2006.

Description of amendment request: The proposed change removes Arkansas Nuclear One, Unit 2 reactor coolant system (RCS) structural integrity requirements contained in Technical Specification (TS) 3.4.10.1. The proposed change is consistent with NUREG-1432, “Standard Technical Specifications—Combustion Engineering Plants,” Revision 3.1.

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 to remove the RCS structural integrity controls from the TSs does not impact any mitigation equipment or the ability of the RCS pressure boundary to fulfill any required safety function. Since no accident mitigation or initiators are impacted by this change, no design basis accidents are affected.

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

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

Response: No.

The proposed change will not alter the plant configuration or change the manner in which the plant is operated. No new failure modes are being introduced by the proposed 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 the margin of safety?

Response: No.

Removal of TS 3.4.10.1 from the TSs does not reduce the controls that are required to maintain the RCS pressure boundary for ASME Code [American Society of Mechanical Engineers' Boiler and Pressure Vessel Code] Class 1, 2, or 3 components. No equipment or RCS safety margins are impacted due to the proposed change.

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

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

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: David Terao.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

Date of amendment request: January 27, 2006.

Description of amendment request: The proposed amendment involves changes to Technical Specifications Section 3/4 9.1, “Boron Concentration,” Section 3/4 9.14, “Spent Fuel Storage,” and Section 3/4 5.5.1, “Fuel Storage Criticality.” The proposed license amendment removes reliance on Boraflex as a neutron absorber in Turkey Point Units 3 and 4 spent fuel pool storage racks. To preclude continued loss of reactivity margin due to the ongoing degradation of Boraflex, the neutron absorbing function currently performed by Boraflex will be replaced by some combination of rod cluster control assemblies, Metamic rack inserts, and administrative controls that require mixing higher reactivity fuel with lower-reactivity fuel.

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

1. Would operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. Operation in accordance with proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendments do not change or modify the fuel, fuel handling processes, spent fuel storage racks, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the spent fuel pool cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents:

a. A fuel handling accident (FHA),

b. A cask drop accident,

c. A fuel mispositioning event,

d. A spent fuel pool boron dilution event,

e. A seismic event, and

f. A loss of spent fuel pool cooling event.

The probability of a FHA is not significantly increased because implementation of the proposed amendment will employ the same equipment and process to handle fuel assemblies that is currently used. Also, tests have confirmed that the Metamic inserts can be installed and removed without damaging the host fuel assemblies. The FHA radiological consequences are not increased because the radiological source term of a single fuel assembly will remain unchanged. Therefore, the proposed amendments do not significantly increase the probability or consequences of a FHA.

The proposed amendments do not increase the probability of dropping a fuel transfer cask because they do not introduce any new heavy loads to the SFP and do not affect heavy load handling processes. Also, the insertion of Metamic rack inserts does not increase the consequences of the cask drop accident because the radiological source term of that accident is developed from a non-mechanistically derived quantity of damaged fuel stored in the spent fuel pool. Therefore, the proposed amendments do not significantly increase the probability or consequences of a cask drop accident.

Operation in accordance with the proposed amendment will not change the probability of a fuel mispositioning event because fuel movement will continue to be controlled by approved fuel handling procedures. These procedures continue to require identification Start Printed Page 27000of the initial and target locations for each fuel assembly that is moved. The consequences of a fuel mispositioning event are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for the worst-case fuel mispositioning event.

Operation in accordance with the proposed amendment will not change the probability of a boron dilution event because the systems and events that could affect spent fuel soluble boron are unchanged. The consequences of a boron dilution event are unchanged because the proposed amendment reduces the soluble boron requirement below the currently required value and the maximum possible water volume displaced by the inserts is an insignificant fraction of the total spent fuel pool water volume.

Operation in accordance with the proposed amendment will not change the probability of a seismic event, which is an Act of God. The consequences of a seismic event are not significantly increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks with Metamic inserts is not appreciably increased. Seismic analyses demonstrate adequate stress levels in the storage racks when inserts are installed.

Operation in accordance with the proposed amendment will not change the probability of a loss of SFP cooling event because the systems and events that could affect SFP cooling are unchanged. The consequences are not significantly increased because there are no changes in the SFP heat load or SFP cooling systems, structures or components. Furthermore, conservative analyses indicate that the current design requirements and criteria continue to be met with the Metamic inserts installed.

Based on the above, it is concluded that the proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Would operation of the facility in accordance with the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

No. Operation in accordance with the proposed amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendments do not change or modify the fuel, fuel handling processes, spent fuel racks, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or the spent fuel pool cooling and cleanup system. The effects of operating with the proposed amendment are listed below. The proposed amendments were evaluated for the potential of each effect to create the possibility of a new or different kind of accident:

a. Addition of inserts to the spent fuel storage racks,

b. New storage patterns,

c. Additional weight from the inserts,

d. Insert movement above spent fuel, and

e. Displacement of fuel pool water by the inserts.

Each insert will be placed between a fuel assembly and the storage cell wall, taking up some of the space available on two sides of the fuel assembly. Tests confirm that the insert can be installed and removed without damaging the fuel assembly. Analyses demonstrate that the presence of the inserts does not adversely affect spent fuel cooling, seismic capability, or subcriticality. The aluminum (alloy 6061) and boron carbide materials of construction have been shown to be compatible with nuclear fuel, storage racks and spent fuel pool environments, and generate no adverse material interactions. Therefore, placing the inserts into the spent fuelpool storage racks can not cause a new or different kind of accident.

Operation with the proposed fuel storage patterns will not create a new or different kind of accident because fuel movement will continue to be controlled by approved fuel handling procedures. These procedures continue to require identification of the initial and target locations for each fuel assembly that is moved. There are no changes in the criteria or design requirements pertaining to spent fuel safety, including subcriticality requirements, and analyses demonstrate that the proposed storage patterns meet these requirements and criteria with adequate margins. Therefore, the proposed storage patterns can not cause a new or different kind of accident.

Operation with the added weight of the Metamic inserts will not create a new or different accident. The net effect of the adding the maximum number of inserts is to add less than one percent to the weight of the loaded racks. Furthermore, the analyses of the racks with Metamic inserts installed demonstrate that the stress levels in the rack modules continue to be considerably less than allowable stress limits. Therefore, the added weight from the inserts can not cause a new or different kind of accident.

Operation with the insert allowed to move above spent fuel will not create a new or different kind of accident. The insert with its handling tool weighs considerably less than the weight of a single fuel assembly. Single fuel assemblies are routinely moved safely over spent fuel assemblies and the same level of safety in design and operation will be maintained when moving the inserts. Furthermore, the effect of a dropped insert to block the top of a storage cell has been evaluated in thermal-hydraulic analyses. Therefore, the movement of inserts can not cause a new or different kind of accident.

Whereas the installed rack inserts will displace a very small fraction of the fuel pool water volume and impose a very small reduction in operator response time to previously-evaluated SFP accidents, the reduction will not promote a new or different kind of accident. Also, displacement of water along two sides of a stored fuel assembly may have some local reduction in the peripheral cooling flow; however, this effect would be small compared to the flow induced through the fuel assembly and would in no way promote a new or different kind of accident.

Based on the above, it is concluded that operation with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Would operation of the facility in accordance with the proposed amendment involve a significant reduction in a margin of safety?

No. Operation of the facility in accordance with the proposed amendment does not significantly reduce the margin of safety. The proposed change was evaluated for its effect on current margins of safety related to criticality, structural integrity, and spent fuel heat removal capability. The margin of safety for subcriticality required by 10 CFR 50.68(b)(4) is unchanged. New criticality analysis confirms that operation in accordance with the proposed amendment continues to meet the required subcriticality margins. Also, the margin of safety for SFP soluble boron concentration is actually increased because new analyses require less soluble boron than is currently required, and much less than the value required by Technical Specifications. The structural evaluations for the racks and spent fuel pool with Metamic inserts installed show that the rack and spent fuel pool are unimpaired by loading combinations during seismic motion, and there is no adverse seismic-induced interaction between the rack and Metamic inserts.

The proposed change does not affect spent fuel heat generation or the spent fuel cooling systems. A conservative analysis indicates that the design basis requirements and criteria for spent fuel cooling continue to be met with the Metamic inserts in place, and displacing coolant. Thermal hydraulic analysis of the local effects of an installed rack insert blocking peripheral flow show a small increase in local water and fuel clad temperatures, but will remain within acceptable limits including no departure from nucleate boiling.

Based on these evaluations, operating the facility with the proposed amendment does not involve a significant reduction in any margin of safety.

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

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

NRC Branch Chief: Michael L. Marshall, Jr.

Nuclear Management Company, LLC, Docket No. 50-306, Prairie Island Nuclear Generating Plant, Unit 2, Goodhue County, Minnesota

Date of amendment request: March 13, 2006.

Description of amendment request: The proposed amendment would involve revision of the surveillance test load in Technical Specification (TS) 3.8.1, “AC Sources—Operating,” Surveillance Requirement (SR) 3.8.1.3. This license amendment request proposes to revise SR 3.8.1.3 to require Start Printed Page 27001testing D5 and D6 monthly at or above 4000 kW to demonstrate TS operability. In addition to the TS required testing, NMC will continue monthly operation at or above 90 percent of the emergency diesel generator (EDG) rated load to assist in early identification of degraded EDG capabilities which could prevent performance of their safety function.

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.

This license amendment request proposes to reduce the Prairie Island Nuclear Generating Plant Unit 2 emergency diesel generator's monthly test loading which demonstrates Technical Specification operability. The proposed test load will continue to assure that both Unit 2 emergency diesel generators have the capacity and the capability to assume the maximum auto-connected loads for Unit 2.

The emergency diesel generators are required to be operable in the event of a design basis accident coincident with a loss of offsite power to mitigate the consequences of the accident. They are also the alternate AC source for a station blackout on the other Prairie Island Nuclear Generating Plant unit. The emergency diesel generators are not accident initiators and therefore this change does not involve a significant increase in the probability of an accident previously evaluated.

The accident analyses assume that at least one safeguards bus is provided with power either from the offsite sources or the emergency diesel generators. The Technical Specification changes proposed in this license amendment request will continue to assure that both Unit 2 emergency diesel generators have the capacity and the capability to assume the maximum auto-connected loads for Unit 2. Thus, the changes proposed in this license amendment request do not involve a significant increase in the consequences of an accident previously evaluated.

The changes proposed in this license amendment do not involve a significant increase the probability or consequences of an accident previously evaluated.

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

Response: No.

This license amendment request proposes to reduce the Prairie Island Nuclear Generating Plant Unit 2 emergency diesel generator's monthly test loading which demonstrates Technical Specification operability. The proposed test load will continue to assure that both Unit 2 emergency diesel generators have the capacity and the capability to assume the maximum auto-connected loads for Unit 2.

The proposed Technical Specification changes do not involve a change in the plant design, system operation, or the use of the emergency diesel generators. The proposed changes allow the emergency diesel generator to be tested at a reduced load which envelopes the required safety function loads and continues to demonstrate the capability and capacity of the emergency diesel generators to perform their required functions. There are no new failure modes or mechanisms created due to testing the emergency diesel generators at the proposed test loading. Testing of the emergency diesel generators at the proposed test loading does not involve any modification in the operational limits or physical design of plant systems. There are no new accident precursors generated due to the proposed test loading.

The Technical Specification changes proposed in this license amendment 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.

This license amendment request proposes to reduce the Prairie Island Nuclear Generating Plant Unit 2 emergency diesel generator's monthly test loading which demonstrates Technical Specification operability. The proposed test load will continue to assure that both Unit 2 emergency diesel generators have the capacity and the capability to assume the maximum auto-connected loads for Unit 2.

The proposed Technical Specification changes will continue to demonstrate that the emergency diesel generators meet the Technical Specification definition of operability, that is, the proposed testing will demonstrate that the emergency diesel generators will perform their safety function and the necessary emergency diesel generator attendant instrumentation, controls, cooling, lubrication and other auxiliary equipment required for the emergency diesel generators to perform their safety function loads are also tested at this loading. The proposed testing will also continue to demonstrate the capability and capacity of the emergency diesel generators to supply the required Unit 2 loss of offsite power coincident with Unit 1 station blackout loads. Since the proposed surveillance testing will continue to demonstrate operability, and the capability and capacity to supply their required Unit 2 loss of offsite power coincident with Unit 1 station blackout loads, the proposed Technical Specification changes do not involve a significant reduction in a margin of safety.

The Technical Specification changes proposed in this license amendment do not involve a significant reduction in a margin of safety.

Based on the above, the Nuclear Management Company concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of “no significant hazards consideration” is justified.

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

Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.

NRC Branch Chief: L. Raghavan.

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

Date of amendment request: February 1, 2006.

Description of amendment request: The proposed amendment would clarify the Technical Specification (TS) testing frequency for the Surveillance Requirements (SRs) in TS 3.1.4, “Control Rod Scram Times.”

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 control rod hydraulic scram insertion system is not an initiator to any accident sequence analyzed in the Final Safety Analysis Report (FSAR). The changes do not involve any physical change to structures, systems, or components (SSCs) and do not alter the method of operation or control of SSCs. The current assumptions in the safety analysis regarding accident initiators and mitigation of accidents (including assumed scram insertion times) are unaffected by these changes. No additional failure modes or mechanisms are being introduced and the likelihood of previously analyzed failures remains unchanged.

Operation in accordance with the proposed Technical Specification (TS) ensures that the control rods and associated scram insertion function remain capable of performing the function as described in the FSAR [Final Safety Analysis Report]. Therefore, the mitigative scram functions will continue to provide the protection assumed by the analysis.

Therefore, this 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 Start Printed Page 27002accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There are no setpoints affected by this change at which protective or mitigative actions are initiated. This change will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the FSAR. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis.

[Therefore, this 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 margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. Operation in accordance with the proposed TS ensures that the control rod scram insertion system remains capable of performing the function as described in the FSAR. Sufficiently rapid insertion of control rods following certain accidents (scram time) will prevent fuel damage, and thereby maintain a margin of safety to fuel damage. No change is being made to the required insertion rate specified in plant Technical Specifications. Clarifying when control rod insertion times must be verified following movement of fuel assemblies, without actually changing the requirement (verification of insertion times will continue to be required whenever work that might impact the rod insertion time is done), does not reduce the margin of safety related to fuel damage.

Therefore, the 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: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Branch Chief: Richard J. Laufer.

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

Date of amendment request: October 7, 2005.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to clarify certain requirements during fuel movement and core alterations. The amendment would make the TSs consistent with the NRC-approved Revision 2 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-51, “Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations,” and NUREG-1433, “Standard Technical Specifications General Electric Plants, BWR [boiling water reactor]/4.”

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 analyzed?

Response: No.

The proposed changes would revise Technical Specifications (TS) 3.6.5.3.1, FRVS [filtration, recirculation and ventilation system] Ventilation System, and 3.6.5.3.2, FRVS Recirculation System, ACTION b from, “* * * containment or operations * * * ” to read “* * * containment and operations * * * ” to be consistent with NUREG-1433, “Standard Technical Specifications General Electric Plants, BWR/4” (STS). Technical Specification 3.7.1.2, Service Water, and 3.8.3.2, Distribution—Shutdown, require the addition of “recently” to modify irradiated fuel consistent with NRC-approved Revision 2 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-51, “Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations.” Technical Specifications 3.8.1.2, A.C. Sources—Shutdown, 3.8.2.2, DC Sources—Shutdown, and 3.8.3.2, Distribution—Shutdown, require that “CORE ALTERATIONS” be added to ACTION a.

The proposed changes associated with the fuel handling accident (FHA) do not involve a change to structures, components, or systems that would affect the probability of an accident previously evaluated in the Hope Creek Updated Final Safety Analysis Report (UFSAR). The FHA for Hope Creek is defined as a drop of a fuel assembly over irradiated assemblies in the reactor core 24 hours after reactor shutdown. 10 CFR 50.67, “Accident Source Term” (AST), was used to evaluate the dose consequences of a postulated accident. The FHA has been analyzed without credit for Secondary Containment; Filtration, Recirculation and Ventilation System (FRVS); and CREF [control room emergency filtration] system. The resultant radiological consequences are within the acceptance criteria set forth in 10 CFR 50.67 and Regulatory Guide (RG) 1.183. This amendment does not alter the methodology or equipment used in fuel handling operations. The equipment hatch, personnel air locks, other containment penetrations, or any component thereof is not an accident initiator. Actual fuel handling operations are not affected by the proposed changes.

Consequently the probability of a previously analyzed FHA is not affected by the proposed amendment. No other accident initiator is affected by the proposed changes.

Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or radiological 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 analyzed?

Response: No.

The proposed changes would revise TS 3.6.5.3.1, FRVS Ventilation System and 3.6.5.3.2, FRVS Recirculation System, ACTION b from, “* * * containment or operations * * * ” to read “* * * containment and operations * * * ” to be consistent with NUREG-1433, Standard Technical Specifications General Electric Plants, BWR/4” (STS). TS 3.7.1.2, Service Water, and 3.8.3.2, Distribution—Shutdown, require the addition of “recently” to modify irradiated fuel consistent with NRC-approved Revision 2 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-51, “Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations.” TS 3.8.1.2 A.C. Sources—Shutdown, 3.8.2.2, D.C. Sources—Shutdown, and 3.8.3.2, Distribution—Shutdown, require that “CORE ALTERATIONS” be added to ACTION a.

The proposed amendment will not create the possibility of a new or different type of accident from any accident previously evaluated because changes to the allowable activity in the primary and secondary systems do not result in changes to the design or operation of these systems. The evaluation of the proposed changes indicates that all design standard and applicable safety criteria limits are met. Equipment important to safety will continue to operate as designed. Component integrity is not challenged. The changes do not result in any event previously deemed incredible being made credible. The changes do not result in more adverse conditions or result in any increase in the challenges to safety systems. The systems affected by the changes are used to mitigate the consequences of a potential accident and would not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed changes would revise TS 3.6.5.3.1, FRVS Ventilation System and 3.6.5.3.2 FRVS Recirculation System, ACTION b from “* * * containment or operations * * * ” to read “* * * containment and operations * * * ” to be consistent with NUREG-1433, “Standard Technical Specifications General Electric Plants, BWR/4” (STS). TS 3.7.1.2, Service Water, and 3.8.3.2, Distribution—Shutdown, require the addition of “recently” to modify irradiated fuel consistent with NRC approved Revision 2 to Technical Specification Task Start Printed Page 27003Force (TSTF) Standard Technical Specification Change Traveler, TSTF-51, “Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations.” TS 3.8.1.2 A.C. Sources—Shutdown, 3.8.2.2 D.C. Sources—Shutdown, and 3.8.3.2 Distribution—Shutdown, require that “CORE ALTERATIONS” be added to ACTION a.

The proposed changes revise the TS operational conditions where specific activities represent situations during which significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis and are established such that the radiological consequences remain at or below the regulatory guidelines. Safety margins and analytical conservatisms are retained to ensure that the analysis adequately bounds all postulated event scenarios. The proposed TS continue to ensure that the total effective dose equivalent (TEDE) for the control room (CR), the exclusion area boundary (EAB), and low population zone (LPZ) boundaries are below the corresponding acceptance criteria specified in 10 CFR 50.67 and RG 1.183.

Therefore, these 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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: Darrell J. Roberts.

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

Date of amendment request: February 23, 2006.

Description of amendment request: The amendment would revise the Operating License Condition 2.C.(6), “Fuel Storage and Handling,” to clarify that the condition does not apply to Nuclear Regulator Commission (NRC)-approved dry spent fuel storage systems. The current condition states no more than a total of three fuel assemblies shall be out of approved shipping containers, fuel assembly storage racks or the reactor at any one time.

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 is a clarification to the Hope Creek operating license to recognize that the dry spent fuel storage system used at the ISFSI [independent spent fuel storage installation] is licensed separately by the NRC under 10 CFR part 72. The change does not affect any SSCs [structure, systems and components] used to operate the reactor or produce electrical power. The change also does not affect SSCs used to shut down the reactor, maintain it in a safe shutdown condition, or mitigate accidents.

The dry storage cask system design is supported by an NRC-approved criticality analysis that demonstrates the system will remain safely subcritical under all normal, off-normal, and credible accident conditions applicable to the dry spent fuel storage system, as defined in the cask CoC holder's 10 CFR part 72 licensing basis. Dry spent fuel storage system loading operations are not addressed in any Part 50 accident as described in Chapter 15 of the HCGS [Hope Creek Generating Station] FSAR [final safety analysis report]. Dry spent fuel storage system loading in the spent fuel pool is governed by procedures that are consistent with the requirements in the HI-STORM 100 System 10 CFR part 72 FSAR. Heavy load handling inside the Part 50 facility associated with cask loading is conducted in accordance with procedures that comply with the site's existing heavy load control program. Because this change does not affect PSEG's [PSEG Nuclear, LLC] heavy load handling procedures and all structures, systems and components used for cask handling will meet the existing commitments to NUREG-0612, a cask drop event remains non-credible as currently described in HCGS FSAR Section 15.7.5.

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 is a clarification to the Hope Creek operating license to recognize that the dry spent fuel storage system is licensed separately by the NRC under 10 CFR part 72. The change does not affect any SSCs used to operate the reactor or produce electrical power. The change also does not affect SSCs used to shut down the reactor, maintain it in a safe shutdown condition, or mitigate accidents.

The dry spent fuel storage system design is supported by an NRC-approved criticality analysis that demonstrates the system will remain safely subcritical under all normal, off-normal, and credible accident conditions, as defined in the cask CoC holder's 10 CFR part 72 licensing basis. Dry spent fuel storage system loading in the spent fuel pool is governed by procedures that are consistent with the requirements in the HI-STORM 100 System 10 CFR 72 FSAR. Heavy load handling inside the Part 50 facility associated with cask loading is conducted in accordance with procedures that comply with the site's existing heavy load control program. Because this change does not affect PSEG's heavy load handling procedures and all structures, systems and components used for cask handling will meet the existing commitments to NUREG-0612, a cask drop event remains non-credible as currently described in HCGS FSAR Section 15.7.5.

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 the margin of safety?

Response: No.

The proposed change is a clarification to the Hope Creek operating license to recognize that dry spent fuel storage systems are licensed separately by the NRC under 10 CFR Part 72. The change does not affect any SSCs used to operate the reactor or produce electrical power. The change also does not affect SSCs used to shut down the reactor, maintain it in a safe shutdown condition, or mitigate accidents.

All safety analyses are consistent with the operations described in the dry spent fuel storage system FSAR and have been previously approved by the NRC as having sufficient safety margins. This change does not affect the dry spent fuel storage system operation procedures or change any normal, off-normal, or accident condition for which the dry spent fuel storage system is designed.

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

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

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

NRC 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: April 17, 2006.

Description of amendment requests: The proposed amendments would delete Section 2.G of the Facility Operating Licenses, which require reporting of violations of the requirements in Sections 2.C(1), 2.C(3), and 2.F of the Facility Operating Licenses.

The NRC staff issued a notice of opportunity for comment in the Federal Register on August 29, 2005 (70 FR 51098), including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated Start Printed Page 27004line item improvement process. The licensee affirmed the applicability of the following NSHC determination in its application dated April 17, 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:

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

Response: No.

The proposed change involves the deletion of a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not significantly increase 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?

Response: No.

The proposed change is administrative in that it deletes a reporting requirement. The change does not add new plant equipment, change existing plant equipment, or affect the operating practices of the facility. Therefore, the 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 change deletes a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not involve a significant reduction in a margin of safety.

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.

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: March 29, 2006.

Description of amendment request: The proposed amendment would revise Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Technical Specifications (TSs) 5.5, “Programs and Manuals,” TS 5.6, “Reporting Requirements,” and TS Bases for LCO [Limiting Condition for Operation] 3.6.1, “Containment,” to reflect the latest requirements for tendon surveillance.

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

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

The proposed change replaces the current TS requirement to implement a Containment Tendon Surveillance Program based on Regulatory Guide 1.35, Rev. 2, with a Containment Inspection Program Plan that complies with the current requirements of 10 CFR 50.55a. This regulation requires licensees to implement a Containment Inspection Program Plan in compliance with the 1992 Edition with the 1992 Addenda of Subsection IWE, “Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants,” and with Subsection IWL, “Requirements for Class CC Concrete Components of Light-Water Cooled Plants,” of Section XI, Division 1, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) with additional modifications and limitations as stated in 10 CFR 50.55a(b)(2)(ix). [Southern Nuclear Operating Company, Inc.] SNC has implemented a Containment Inspection Program Plan that complies with the regulatory requirements. This proposed TS amendment is requested to update the TS to the latest 10 CFR 50.55a regulatory requirements.

In addition, reporting requirements that are redundant to existing regulations are deleted, minor editorial changes are made, and the applicability of SR 3.0.2 to the tendon surveillance program is deleted since surveillance frequencies and associated extensions are specified in ASME Section XI, Subsection IWL.

By complying with the regulatory requirements described in 10 CFR 50.55a, the probability of a loss of containment structural integrity is maintained as low as reasonably achievable. Maintaining containment structural integrity as described in the revised Containment Inspection Program Plan does not impact the operation of the reactor coolant system (RCS), containment spray (CS) system, or emergency core cooling system (ECCS). The Containment Inspection Program ensures that the containment will function as designed to provide an acceptable barrier to release of radioactive materials to the environment. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

The proposed change does not impact any accident initiators or analyzed events, nor does it impact the types or amounts of radioactive effluent that may be released offsite. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Maintaining containment structural integrity does not impact the operation of the RCS, CS system, or ECCS. The proposed change does not involve a modification to the physical configuration of the plant or a change in the methods governing normal plant operation. The proposed change does not introduce a new accident initiator, accident precursor, or malfunction mechanism. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed license amendment does not involve a significant reduction in a margin of safety.

By complying with the regulatory requirements described in 10 CFR 50.55a, the probability of a loss of containment structural integrity is maintained as low as reasonably achievable. The Containment Inspection Program Plan ensures that the containment will function as designed to provide an acceptable barrier to release of radioactive materials to the environment. The proposed change does not adversely affect plant operation or existing safety analyses. 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: Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: March 28, 2006.

Description of amendment request: The amendment would delete references to specific isolation valves in the chemical and volume control system (CVCS) and to modify notes to allow (1) an exception for decontamination activities and (2) an exception for CVCS resin vessel operation. These are changes to Technical Specifications (TSs) 3.3.9, “Boron Dilution Mitigation System (BDMS),” and 3.9.2, “Unborated Water Source Isolation Valves.”

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 do not involve a significant increase in the probability or Start Printed Page 27005consequences of an inadvertent boron dilution accident by isolating the CVCS resin vessels in MODE 6 or by isolating the purge line for detector SJRE001 during flushing activities in MODE 6. By recognizing these potential [boron] dilution sources and by making TS 3.3.9 and TS 3.9.2 more generic for consideration of all potential [boron] dilution sources, plant administrative controls are revised such that the plant is put in a safer condition than before. Specific isolation valves are removed from TS 3.3.9 and TS 3.9.2. They are relocated from the [Technical] Specifications to the appropriate TS Bases. This is an administrative only change and is consistent with the [Improved] Standard Technical Specifications, NUREG-1431. [The Wolf Creek Technical Specifications are based on NUREG-1431.] Allowing a [boron] dilution source path to be unisolated under administrative controls, described in TS Bases 3.9.1 during refueling decontamination activities, is acceptable as allowed by Amendment [No.] 97 to the Callaway Operating License and does not involve a significant increase in the probability or consequences of an inadvertent boron dilution accident. Allowing an exception for CVCS resin vessel operation is acceptable because chemistry controls may require some CVCS resin vessels to be configured with resin intended for boron dilution. Plant conditions may warrant their use. As allowed by the LCO [limiting condition for operation] Note, these vessels may be unisolated under administrative controls. The administrative controls ensure that the resin vessels are not [boron] dilution sources [for the reactor coolant system (RCS)]. These changes do not involve a significant increase in the probability or consequences of an inadvertent boron dilution accident.

The proposed changes do not involve a significant increase in the probability or consequences of an inadvertent boron dilution accident by requiring the isolation of all unborated water source isolation valves in higher plant modes when both trains of BDMS are inoperable or when a condition of no reactor coolant loop in operation exists. Proposed TS 3.3.9 Required Actions [B.3.1, B.3.2, C.1 and C.2] are generic and remain consistent with the plant accident analyses. Allowing exceptions for CVCS resin vessel operation is acceptable because chemistry controls may require some CVCS resin vessels to be configured with resin intended for boron dilution. Plant conditions may warrant their use. As allowed by exception Notes, these vessels may be unisolated under administrative controls. The administrative controls ensure that the resin vessels are not [boron] dilution sources.

Therefore, 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 do not create the possibility of a new or different kind of accident. Although other potential [boron] dilution sources are identified for administrative control[s], the evaluation of a MODE 6 [boron] dilution event remains unchanged. Isolating the CVCS resin vessels or isolating the purge line for detector SJRE001 during flushing activities in MODE 6 and making TS 3.3.9 and TS 3.9.2 more generic does not impact the operability of any safety related equipment required for plant operation. No new equipment will be added and no new limiting single failures are created. The plant will continue to be operated within the envelope of the existing safety analysis. In addition[,] specific isolation valves are removed from TS 3.3.9 and TS 3.9.2. They are relocated from the [Technical] Specifications to the appropriate TS Bases. This is an administrative only change and is consistent with the [Improved] Standard Technical Specifications, NUREG-1431. Allowing a [boron] dilution source path to be unisolated under administrative controls, described in TS Bases 3.9.1 during refueling decontamination activities, is acceptable as allowed by Amendment [No.] 97 to the Callaway Operating License and does not create the possibility of a new or different kind of inadvertent boron dilution accident. Allowing an exception for CVCS resin vessel operation is acceptable because chemistry controls may require some CVCS resin vessels to be reconfigured with resin intended for boron dilution. Plant conditions may warrant their use. As allowed by the LCO Note these vessels may be unisolated under administrative controls. The administrative controls ensure that the resin vessels are not [boron] dilution sources. These changes do not create the possibility of a new or different kind of accident from an inadvertent boron dilution accident previously evaluated.

Requiring the isolation of unborated water source isolation valves in higher plant modes when both trains of BDMS are inoperable or when a condition of no RCS loop in operation exists, does not create the possibility of a new or different kind of inadvertent boron dilution accident. Proposed TS 3.3.9 is generic and remains consistent with the plant accident analyses. Allowing exceptions for CVCS resin vessel operation is acceptable because chemistry controls may require some CVCS resin vessels to be configured with resin intended for boron dilution. Plant conditions may warrant their use. As allowed by exception Notes, these vessels may be unisolated under administrative controls. The administrative controls ensure that the resin vessels are not [boron] dilution sources.

Therefore, the proposed changes do not create a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes do not reduce the margin of safety. Although other potential [boron] dilution sources are identified for administrative control[s] and TS 3.3.9 and TS 3.9.2 are made generic for consideration of all potential [boron] dilution sources, the evaluated margin of safety for a [boron] dilution event in MODE 6 remains the same. Recognition of other potential [boron] dilution sources, isolation of the CVCS resin vessels and the purge line for detector SJRE001 during flushing activities in MODE 6, places the plant in a safer condition than before. In addition[,] specific isolation valves are removed from TS 3.3.9 and TS 3.9.2. They are relocated from the [Technical] Specifications to the appropriate TS Bases. This is an administrative only change and is consistent with the [Improved] Standard Technical Specifications, NUREG-1431. Finally, allowing a [boron] dilution source path to be unisolated under administrative controls, described in TS Bases 3.9.1 during refueling decontamination activities, is acceptable under Amendment [No.] 97 to the Callaway Operating License and does not involve a significant reduction in a margin of safety [ * * * ]. Allowing an exception for CVCS resin vessel operation is acceptable because chemistry controls may require some CVCS resin vessels to be configured with resin intended for boron dilution. Plant conditions may warrant their use. As allowed by the LCO Note these vessels may be unisolated under administrative controls. The administrative controls ensure that the resin vessels are not [boron] dilution sources. This change does not involve a significant reduction in a margin of safety [ * * * ].

Requiring the isolation of all unborated water source isolation valves in higher plant modes when both trains of BDMS are inoperable or when no reactor coolant loop is in operation does not involve a significant reduction in the margin of safety. The changes to the [Technical] Specifications make it generic and [remain] consistent with the plant accident analyses. Allowing exceptions for CVCS resin vessel operation is acceptable because chemistry controls may require some CVCS resin vessels to be configured with resin intended for boron dilution. Plant conditions may warrant their use. As allowed by these exception Notes, these vessels may be unisolated under administrative controls. The administrative controls ensure that the resin vessels are not [boron] dilution sources.

Therefore, the proposed changes do not involve a significant reduction in [a] margin of safety.

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

Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: David Terao.

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

Date of amendment request: March 28, 2006.

Description of amendment request: The amendment would revise Technical Specification 5.0, “Administrative Start Printed Page 27006Controls,” by changing position titles and department names. The amendment would not change any specific responsibilities, job functions, organizational commitments, or qualification requirements of plant personnel.

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 do not affect accident initiators or assumptions. The radiological consequences of accidents previously evaluated remain unchanged. These changes involve administrative changes concerning designations for position titles and department names. The changes do not affect responsibilities, functions, organizational commitments, or the qualification requirements of plant personnel.

Therefore, 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 are administrative in nature. The overall operating philosophy of [the] Callaway Plant is unchanged. As such, there are no hardware changes nor are there any changes in the method by which any safety-related plant system performs its safety function. This amendment will not affect the normal method of plant operation or change any operating parameters. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this amendment. There will be no adverse effects or challenges imposed on any safety-related system as a result of this amendment.

Therefore, the proposed changes do not create a new or different kind of accident from any accident previously evaluated.

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

Response: No.

There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. The changes do not involve any change in overall organizational commitments. The changes to personnel titles and department designations are administrative and will not reduce any margin of safety.

Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

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

Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

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 Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear Station, Unit 2, York County, South Carolina

Date of application for amendments: December 19, 2005, as supplemented on February 2 and 28, 2006.

Brief description of amendments: The amendment made a one-time change to the Technical Specifications regarding the required steam generator (SG) tube repair criteria for Catawba Unit 2 during refueling outage 14 and operating cycle 15. In addition, the proposed amendment added a license condition that requires a reduction in the allowable normal operating primary-to-secondary leakage rate from 150 gallons-per-day to 75 gallons-per-day through any one SG and from 600 gallons-per-day to 300 gallons-per-day through all SGs. The proposed license condition will be applicable only for the duration of Catawba Unit 2 cycle 15 operation.

Date of issuance: March 31, 2006.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance March 31, 2006.

Amendment No.: 224.

Renewed Facility Operating License No. NPF-52: Amendments revised the Technical Specifications and the license.

Date of initial notice in Federal Register: February 22, 2006 (71 FR 9169).

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

No significant hazards consideration comments received: No. Start Printed Page 27007

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

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

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

Duke Energy Corporation, Docket No. 72-004, Oconee Independent Spent Fuel Storage Installation, Oconee County, South Carolina

Date of application for amendments: August 5, 2005, as supplemented by letters dated November 28 and December 14, 2005, and February 6, 2006.

Brief description of amendments: The amendments revised the operating licenses approving the indirect transfer of the Renewed Facility Operating Licenses for Catawba Nuclear Station, Units 1 and 2, McGuire Nuclear Station, Units 1 and 2, and Oconee Nuclear Station, Units 1, 2, and 3, and the Materials License for Oconee Independent Spent Fuel Storage Installation from Duke Energy Corporation to a new holding company, to be named Duke Energy Corporation, in connection with a proposed corporate restructuring and merger involving Cinergy Corporation.

Date of issuance: April 1, 2006.

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

Amendment Nos.: 229, 225, 232, 214, 349, 351, 349 and 8 respectively.

Renewed Facility Operating License Nos. NPF-35 , NPF-52, NPF-9, NPF-17, DPR-38, DPR-47, DPR-55, and SNM-2503: Amendments revised the Operating Licenses.

Date of initial notice in Federal Register: December 30, 2005 (70 FR 77428).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 7, 2006 (ML060250498).

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of application for amendment: May 19, 2004.

Brief description of amendment: The change revises Technical Specification (TS) 3.8.1, “AC Sources—Operating,” to permit a longer completion time for the Division 1 and Division 2 diesel generators (DGs). This is a risk-informed TS change that would extend the DG completion time from 72 hours (the current limit) to 14 days.

Date of issuance: April 14, 2006.

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

Amendment No.: 197.

Facility Operating License No. NPF-21: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 22, 2004 (69 FR 34699).

The September 1, 2005, January 9, February 23, and March 20, 2006, supplemental letters and March 30, 2006, e-mail 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 no significant hazards considerations determination.

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 2, 2004, as supplemented by letters dated August 9, 2005, December 29, 2005 and March 22, 2006.

Brief description of amendment: The amendment allows continued plant operation with a single recirculation loop operation at Pilgrim.

Date of issuance: April 12, 2006.

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

Amendment No.: 219.

Facility Operating License No. DPR-35: The amendment revised the Facility Operating License, Technical Specifications and Surveillance Requirements.

Date of initial notice in Federal Register: December 21, 2004 (69 FR 76490).

The supplements dated August 9, 2005, December 29, 2005 and March 22, 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 consideration determination.

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 24, 2005.

Brief description of amendment: The amendment deletes the main steam isolation valve twice per week partial stroke testing surveillance specified in Technical Specification 4.7.A.2.b.1.c.

Date of issuance: April 13, 2006.

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

Amendment No.: 220.

Facility Operating License No. DPR-35: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 16, 2005 (70 FR 48205).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 13, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 24, 2005, as supplemented by letter dated December 6, 2005.

Brief description of amendment: The amendment revises the Technical Specifications allowances for bypassing the rod worth minimizer.

Date of issuance: April 13, 2006.

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

Amendment No.: 221.

Facility Operating License No. DPR-35: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 30, 2005 (70 FR 51380).

The supplement dated December 6, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 13, 2006.

No significant hazards consideration comments received: No. Start Printed Page 27008

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

Date of application for amendment: April 20, 2005.

Brief description of amendment: The changes revised the Technical Specifications (TSs) to replace plant-specific position titles with generic position titles. Also, the changes deleted TS 6.7, “Safety Limit Violations or Protective Limit Violation,” and included a change to TS 2.1.2, “Reactor Core,” associated with the deletion of TS 6.7. Additionally, the changes relocated to the Davis-Besse Nuclear Power Station Updated Safety Analysis Report the Process Control Program requirements from TS 6.8, “Procedures and Programs,” and from TS 6.14, “Process Control Program (PCP).” Associated with this change, TS Definition 1.30, “Process Control Program,” was deleted. Also, TS 6.15, “Offsite Dose Calculation Manual (ODCM),” was modified to eliminate the requirement that changes to the ODCM be reviewed and accepted by the Plant Operations Review Committee (PORC). These changes to administrative requirements also eliminated the need to propose additional changes in the future to plant-specific position/organizational titles. The changes are consistent with NUREG-1430, “Standard Technical Specifications—Babcock and Wilcox Plants,” Revision 3, dated June 2004. Lastly, the changes revised in the TSs the title “Industrial Security Plan” to “Physical Security Plan.”

Date of issuance: February 7, 2006.

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

Amendment No.: 272.

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

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

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

No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: January 6, 2005, as supplemented October 14, 2005, and February 13, 2006.

Brief description of amendment: The amendment revises Technical Specification (TS) Section 3/4.4.5, “Steam Generators,” to allow repair of steam generator tubes by installing Westinghouse Alloy 800 leak limiting sleeves.

Date of Issuance: April 18, 2006.

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

Amendment No.: 144.

Renewed Facility Operating License No. NPF-16: Amendment revised the TS.

Date of initial notice in Federal Register: March 1, 2005 (70 FR 9993). The October 14, 2005, and February 13, 2006, supplements did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 18, 2006.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendments request: June 1, 2005, as supplemented on February 13, 2006.

Brief Description of amendments: The amendments revise Technical Specification (TS) Section 5.5.6, “Pre-Stressed Concrete Containment Tendon Surveillance Program,” for consistency with the requirements of 10 CFR 50.55a(g)(4) for components classified as Code Class CC. The amendments also delete the provisions of Surveillance Requirement 3.0.2 from this TS and delete the reporting requirements in TS 5.6.9, “Tendon Surveillance Report.”

Date of issuance: April 14, 2006.

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

Amendment Nos.: 172 and 165.

Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: June 21, 2005 (70 FR 35739). The February 13, 2006, supplemental letter provided clarifying information that did not change the June 1, 2005, application and the initial proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 29, 2005.

Brief description of amendments: The proposed amendments revised the technical specification testing frequency for the surveillance requirement 3.1.4.2, control rod scram time testing, from 120 days cumulative operation in MODE 1 to 200 days cumulative operation in MODE 1.

Date of issuance: January 9, 2006.

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

Amendment Nos.: 295 and 253.

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

Date of initial notice in Federal Register: September 27, 2005 (70 FR 56504).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 9, 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 amendment request: January 24, 2005.

Brief description of amendments: The requested amendments revise Technical Specification (TS) 3.7.5, “Auxiliary Feedwater (AFW) System.” The change would add a Note to surveillance requirements (SRs) 3.7.5.1, 3.7.5.3, and 3.7.5.4 that states, “AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.”

Date of issuance: April 24, 2006.

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

Amendment Nos.: 126 and 126.

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

Date of initial notice in Federal Register: November 8, 2005 (70 FR 67753).

No significant hazards consideration comments received: No. Start Printed Page 27009

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 Start Printed Page 27010statement 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.

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).

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook Nuclear Plant, Unit 2 (DCCNP-2), Berrien County, Michigan

Date of amendment request: April 10, 2006, as supplemented on April 12, and 13 (two letters), 2006.

Description of amendment request: The amendment revised Surveillance Requirement 3.8.1.11 of the DCCNP-2 Technical Specifications, raising the diesel generator load rejection voltage test limit from 5000 volts to 5350 volts.

Date of issuance: April 13, 2006.

Effective date: April 13, 2006.

Amendment No.: 276.

Facility Operating License No. DPR-74: Amendment revises the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): No. The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated April 13, 2006.

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

NRC Branch Chief: L. Raghavan.

Start Signature

Dated at Rockville, Maryland, this 1st day of May 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. 06-4243 Filed 5-8-06; 8:45 am]

BILLING CODE 7590-01-P