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

Table of Contents

I. Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from August 28, 2008 to September 10, 2008. The last biweekly notice was published on September 9, 2008 (73 FR 52412).

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.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rulemaking, Directives and Editing Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 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, person(s) 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 via electronic submission through the NRC E-Filing system 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 person(s) 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'sproperty, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for hearing or a petition for leave to intervene must be filed in accordance with the NRC E-Filing rule, which the NRC promulgated in August 28, 2007 (72 FR 49139). The E-Filing process requires participants to submit and serve documents over the internet or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek a waiver in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms Viewer TM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms Viewer [TM] is free and is available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.

Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The EIE system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically may seek assistance through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line, which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, Monday through Friday. The help line number is (800) 397-4209 or locally, (301) 415-4737.

Participants who believe that they have a good cause for not submitting documents electronically must file a motion, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted 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; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service.

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later than 11:59 p.m. Eastern Time on the due date.

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a Presiding Officer. Participants are requested not to includepersonal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

For further details with respect to this amendment 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.

Carolina Power Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of amendments request: July 7, 2008

Description of amendments request: The proposed change would revise Surveillance Requirement (SR) 3.6.1.6.1 to add a new requirement to verify that each vacuum breaker is closed within 6 hours following an operation that causes any of the vacuum breakers to open and revises SR 3.6.1.6.2 by removing the requirement to perform functional testing of each vacuum breaker within 12 hours following an operation that causes any of the vacuum breakers to open.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR Part 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 does not involve physical changes to any plant structure, system, or component. The suppression chamber-to-drywell vacuum breakers only provide an accident mitigation function. As such, the probability of occurrence for a previously analyzed accident is not impacted by the change to the surveillance frequency for these components.

The consequences of a previously analyzed accident are dependent on the initial conditions assumed for the analysis, the behavior of the fuel during the analyzed accident, the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. No physical change to suppression chamber-to-drywell vacuum breakers is being made as a result of the proposed change, nor does the change alter the manner in which the vacuum breakers operate during an accident. As a result, no new failure modes of the suppression chamber-to-drywell vacuum breakers are being introduced. The surveillance requirements for the suppression chamber-to-drywell vacuum breakers will continue to ensure testing of the suppression chamber-to-drywell vacuum breakers following plant transients involving the discharge of steam to the suppression chamber from the SRVs, and such testing will continue to provide assurance that the vacuum breakers are able to perform their design function. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.

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

Response: No.

The proposed change to the surveillance requirements for the suppression chamber-to-drywell vacuum breakers does not involve any physical alteration of plant systems, structures, or components. No new or different equipment is being installed. No installed equipment is being operated in a different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions. As a result no new failure modes are being introduced. Therefore, the proposed change to the surveillance requirements for the suppression chamber-to-drywell vacuum breakers does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

Response: No.

The proposed change revises Surveillance Requirement 3.6.1.6.1 to add a new requirement to verify each vacuum breaker is closed within 6 hours following an operation that causes any of the vacuum breakers to open and revises Surveillance Requirement 3.6.1.6.2 by removing the requirement to perform functional testing of each vacuum breaker within 12 hours following an operation that causes any of the vacuum breakers to open. The operability and functional characteristics of the suppression chamber-to-drywell vacuum breakers remains unchanged. The margin of safety is established through the design of the plant structures, systems, and components, through the parameters within which the plant is operated, through the establishment of the setpoints for the actuation of equipment relied upon to respond to an event, and through margins contained within the safety analyses. The proposed change to the surveillance requirements for the suppression chamber-to-drywell vacuum breakers does not impact the condition or performance of structures, systems, setpoints, and components relied upon for accident mitigation. The proposed change to Surveillance Requirements 3.6.1.6.1 and 3.6.1.6.2 will avoid unnecessary cycling and wear of the vacuum breaker test actuation mechanisms, will improve the reliability of the vacuum breakers, and will minimize the potential for a plant shut down due to a problem with a vacuum breaker test actuating mechanism from excessive wear. The proposed change does not impact any safety analysis assumptions or results. Therefore, the proposed change does not result in 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 proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Thomas H. Boyce.

Dominion Nuclear Connecticut, Inc. Docket Nos. 50-245, 50-336, and 50-423, Millstone Power Station, Units 1, 2, and 3, New London County, Connecticut

Date of amendment request: August 21, 2008.

Description of amendment request: The proposed amendment removes references to and limits imposed by Nuclear Regulatory Commission Generic Letter (GL) 82-12, “Nuclear Power Plant Staff Working Hours,” from the subject plants” technical specifications (TS). The guidelines have been superseded by the requirements of Title 10 of the Code of Federal Regulations, Part 26 (10 CFR 26), Subpart I, “Managing Fatigue.”

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 removal of references to GL 82-12 will not remove the requirement to control work hours and manage fatigue. Removal of TS references to GL 82-12 will be performed concurrently with the implementation of the more conservative 10 CFR 26, Subpart I, requirements.

The proposed changes do not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not impact the initiators or assumptions of analyzed events, nor do they impact the mitigation of accidents or transient events.

Because these new requirements are administrative in nature and further, are more conservative with respect to work hour controls and fatigue management, the proposed change will not significantly increase the probability or consequence 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 remove references to GL 82-12 from TS consistent with the recently revised Subpart I to 10 CFR 26. These regulations are more restrictive than the current guidance and would add conservatism to work hour controls and fatigue management. Work hours will continue to be controlled in accordance with NRC requirements. The new rule continues to allow for deviations from controls to mitigate or prevent a condition adverse to safety or necessary to maintain the security of the facility. This ensures that the new rule will not restrict work hours at the expense of the health and safety of the public as well as plant personnel.

The proposed changes do not alter plant configuration, require that new plant equipment be installed, alter assumptions made about accidents previously evaluated, add any initiators, or impact the function of plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected.

Because the proposed changes do not remove the station's requirement to control work hours and increases the conservatism of work hour controls by changing administrative scheduling requirements, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

Compliance with the new rule adds conservatism to existing fatigue management and contributes to the margin of safety. Deletion of references to GL 82-12 in the TS is administrative in nature since fatigue management is controlled through the new rule. MPS1, MPS2 and MPS3 will continue their fitness-for-duty and behavioral observation programs, both of which will be strengthened by compliance with the new rule. The proposed changes add conservatism to fatigue management and contribute to the margin of safety.

The proposed changes do not involve any physical changes to plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting conditions of operation, or design parameters for any SSC.

The proposed changes do not impact any safety analysis assumptions and do not involve a change in initial conditions, system response times, or other parameters affecting an accident analysis.

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 requested amendments involve no significant hazards consideration.

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.

NRC Branch Chief: Harold K. Chernoff.

Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station, Unit1, Oconee County, South Carolina

Date of amendment request: June 26, 2008.

Description of amendment request: The proposed amendment would result in a revision of the current licensing basis (LB) in regard to high-energy line break (HELB) events occurring outside of containment for Oconee Nuclear Station, Unit 1 (ONS-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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Justification: The ONS-1 changes proposed in this LAR [license amendment request] include revisions to the current HELB methodology and mitigation strategy as documented in a new HELB report. This report provides the completed analysis for ONS HELBs including the descriptions of the station modifications that have been or will be made as a result of this comprehensive HELB reanalysis.

The modifications associated with the revised HELB LB will be designed and installed in accordance with applicable quality standards such that the likelihood of failure of new or modified SSCs will not initiate failures, malfunctions, or inadvertent operations of existing accident mitigating SSCs [structures, systems, and components], such as the KHUs [Keowee hydro units], SSF [standby shutdown facility], HPI [high-pressure injection], or the Central Tie Switchyard 100 kV alternate power systems. For Turbine Building HELBs that could adversely affect equipment needed to stabilize and cooldown the units, the addition of the PSW [protected service water] System provides added assurances that safe shutdown can be readily established and maintained beyond the 72-hour SSF mission time.

In conclusion, the changes will collectively enhance the station's overall design, safety, and risk margin; therefore, the proposed change does not involve a significant increase in the probability or consequence 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.

Justification: The proposed modifications address potential adverse consequences from a HELB outside of containment. These modifications will be designed and installed in compliance with applicable quality standards such that there are reasonable assurances that they will neither introduce nor cause new failure mechanisms, malfunctions or accident initiators not already considered in the current HELB design and licensing basis.

The overall effect of the changes to the HELB LB is considered an enhancement to the station's ability to achieve safe and cold shut down following a damaging HELB; therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

Justification: The revised HELB LB will collectively enhance the station's overall design, safety, risk margin, and the station's ability to mitigate a HELB event; therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Duke concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significance 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 request involves no significant hazards consideration.

Attorney for licensee: Ms. Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, EC07H,Charlotte, NC 28202.

NRC Branch Chief: Melanie C. Wong.

Duke Energy Carolinas, LLC, 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 26, 2008.

Description of amendment request: The proposed amendments would resultin a revision to portions of the Updated Final Safety Analysis Report (UFSAR) regarding the tornado licensing basis (LB).

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:

(4) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Justification: Although a tornado does not constitute a previously-evaluated UFSAR Chapter 15 design basis accident or transient as described in 10 CFR 50.36(c)(2), it is a design basis criterion that is required to be considered in plant equipment design. The possibility of a tornado striking the ONS is appropriately considered in the UFSAR and Duke has concluded that the proposed changes do not increase the possibility that a damaging tornado will strike the site or increase the consequences from a damaging tornado.

The modifications associated with the revised tornado LB will be designed and installed such that failures in these new or modified SSCs [structures, systems, and components will not initiate failures or inadvertent operations of existing ONS accident mitigating SSCs, such as the KHUs [Keowee hydro units], SSF [standby shutdown facility], or HPI [high-pressure injection] systems. The use of the NRC-approved TORMIS methodology confirmed that the risk from missile damage was acceptably low to vulnerable areas of the SSF structures and other SSCs required for SSD [safe shutdown]. As a result, there is reasonable assurance that a tornado missile will not prohibit the SSF system from fulfilling its tornado LB or other functions.

Also, there are additional electrical power sources available which provide increased assurance that systems used to transition the units to SSD can be readily powered following a damaging tornado. The PSW [protected service water] System will provide additional assurance that SSD can be established and maintained.

Overall, the changes proposed will increase assurance that potential challenges to the integrity of the RCS due to the effects of a damaging tornado will not result in a radioactive release to the environment. In conclusion, the changes will collectively enhance the station's overall design, safety, and risk margin; therefore, the probability or consequences of accidents previously evaluated are not significantly increased.

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

Response: No.

Justification: Although only the SSF is credited for establishing and maintaining SSDHR [secondary side decay heat removal] and RCMU [reactor coolant makeup] during the first 72 hours following a damaging tornado, there are two relatively independent, diverse and redundant systems capable of safely shutting down all three units in the revised LB (SSF and PSW). Other modifications improve the ability of the SSF and PSW systems to perform their functions following a damaging tornado. The modifications will be designed and installed such that they will not introduce new failure mechanisms, malfunctions or accident initiators not already considered in the design and LB.

In conclusion, the changes to the tornado LB will not degrade existing plant systems and will significantly enhance the station's ability to achieve SSD following a damaging tornado. The design and installation of the PSW system will be such that there is reasonable assurance that the system, including new power paths, will not contribute to the possibility of new or different kind of accident from any accident previously evaluated.

(6) Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Justification: The revised tornado LB will collectively enhance the station's overall design, safety, and risk margin; 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: Ms. Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, EC07H,Charlotte, NC 28202.

NRC Branch Chief: Melanie C. Wong.

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

Date of amendment request: June 26, 2008, as supplemented by letters dated August 4 and August 26, 2008.

Description of amendment request: The proposed amendments would make changes to the Technical Specifications that are conforming or related to a change in fuel type from Westinghouse 0.400-inch OD Vantage+ fuel to Westinghouse 0.422-inch OD Vantage+ 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. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The requested amendment is related to a change in the reload fuel design. The design criteria for the reload fuel are consistent with those for the existing fuel and ensure that the reload fuel is compatible on the basis of coolant flow and neutronic characteristics, as well as DNB and peak cladding temperature requirements. The reload fuel design also ensures mechanical compatibility with the existing fuel, reactor core, control rods, steam supply system, and fuel handling tools and system.

The reactor fuel and its analysis are not accident initiators. Therefore, the change in reload fuel design does not affect accident or transient initiation.

The minimum boron accumulator concentration is also not an accident initiator. The proposed change to the minimum accumulator boron concentration Technical Specification limit ensures that the plant will continue to operate in a manner that provides acceptable levels of protection for health and safety of the public. Further, all design basis accidents and transients affected by the fuel upgrade were re-analyzed or evaluated using representative core designs and the results for each fuel type show all acceptance criteria will continue to be met.

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

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

Response: No.

Use of the 422V+ fuel is consistent with current plant design bases and does not adversely affect any fission product barrier, nor does it alter the safety function of safety significant systems, structures and components or their roles in accident prevention or mitigation. The operational characteristics of 422V+ fuel are bounded by the safety analyses * * *. The 422V+ fuel design performs within existing fuel design limits.

The proposed change to the minimum accumulator boron concentration Technical Specification limit ensures that the plant will continue to operate in a manner that provides acceptable levels of protection for health and safety of the public. Further, all design basis accidents and transients affected by the fuel upgrade were re-analyzed or evaluated using representative core designs and the results for each fuel type show all acceptance criteria will continue to be met.

No equipment additions or modifications are included with the proposed change.

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

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

Response: No.

The proposed changes do not alter the manner in which applicable design basis limits are determined, nor do they result in exceeding existing design basis limits. Thus, all licensed safety margins are maintained.

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 requests involve no significant hazards consideration.

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: Lois M. James.

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

Date of amendment request: June 27, 2008.

Description of amendment request: These proposed changes consist of Proposed Change Number 583 (PCN-583) and are in support of the replacement of the steam generators (SGs) at SONGS Units 2 and 3. The proposed changes reflect revised SG inspection and repair requirements, and revised peak containment post-accident pressure resulting from installation of the replacement SGs.

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

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

Response: No.

The proposed changes will reflect installation of Replacement Steam Generators (RSGs) at San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. The proposed changes involve revising the Steam Generator (SG) tube inspection and repair [requirements] and revising the peak containment post-accident pressure.

The proposed change to revise the SG tube inspection and repair [requirements] affect Technical Specifications (TSs) 3.4.17, “Steam Generator (SG) Tube Integrity,” 5.5.2.11, “Steam Generator (SG) Program,” and 5.7.2.c, “Special Reports.” The proposed TS 3.4.17, 5.5.2.11, and 5.7.2.c revisions remove the repair method (sleeving), and Alternate Repair Criteria (ARC). The revisions replace the 44% tube repair criterion applicable to the original SGs, with a 35% (preliminary) tube repair criterion applicable to the RSGs. The revisions replace inspection requirements applicable to the tubing material of the original SGs with inspection requirements applicable to the tubing material of the RSGs, thus maintaining consistency with applicable material-specific regulatory guidance (TSTF-449, Revision 4). Overall, these revisions will ensure that all RSG tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 35% (preliminary) of the nominal tube wall thickness will be plugged as required by revised TS 5.5.2.11.c.1.

The TS 5.5.2.11.b SG structural integrity, accident induced leakage, and operational leakage performance criteria are unchanged and will continue to be met for the RSGs. Meeting the SG performance criteria provides reasonable assurance that the SG tubing will remain capable of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident.

The proposed change to the SG tube inspection and repair [requirements] will not affect the probability of any accident initiators. There will be no degradation in the performance of, or an increase in the number of challenges imposed on, safety-related equipment assumed to function during an accident. There will be no change to accident mitigation performance. The proposed change will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report (UFSAR).

The proposed change to the peak containment post-accident pressure will revise TS 5.5.2.15, “Containment Leakage Rate Testing Program,” by changing the stated values for peak containment internal pressure for the design-basis Loss-of-Coolant Accident (LOCA) and Main Steam Line Break (MSLB) accidents. The current LOCA value of 45.9 psig would be changed to 48.0 psig and the current MSLB value of 56.5 psig would be changed to 51.5 psig.

The proposed change does not affect the probability of occurrence of an accident previously evaluated because it relates solely to the consequences of hypothesized accidents given that the accident has already occurred.

The proposed change increases the calculated peak containment internal pressure for the LOCA events from 45.9 psig to 48.0 psig. The revised post-LOCA peak containment pressure is bounded by the existing and revised post-MSLB peak containment pressure and the containment design pressure of 60 psig. Despite the increase in the post-LOCA peak containment pressure, any post-accident containment leakage will still be limited to less than 0.1% containment air volume per day, consistent with current TS 5.5.2.15. Therefore, there is no increase in the radiological consequences of a LOCA as a result of the change to the post-LOCA peak containment pressure.

The post-MSLB peak containment pressure decreases from 56.5 psig to 51.5 psig. Thus, the peak containment post-accident pressure is decreased as a result of this change, and there is no resulting increase in the consequences of a previously evaluated accident.

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

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

[Response: No.]

The proposed change to the SG tube inspection and repair [requirements] deletes the repair method (sleeving) and the ARC applicable to the original SGs, and provides repair criteria and inspection requirements applicable to the RSGs. This will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The primary-to-secondary leakage that may be experienced during all plant conditions will be monitored to ensure it remains within current accident analysis assumptions. The proposed change does not adversely affect the method of operation of the SGs or the primary or secondary coolant chemistry controls and does not impact other plant systems or components.

The proposed change to the peak containment post-accident pressure relates to two accidents, LOCA and MSLB, which are already evaluated in the Updated Final Safety Analysis Report (UFSAR).

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

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

Response: No.

For the proposed change to the SG inspection and repair [requirements], the safety function of the SGs is maintained by ensuring the integrity of the tubes. SG tube integrity is a function of the design, environment, and the physical condition of the SG tubes. The proposed change, which deletes the repair method (sleeving) and the ARC applicable to the original SGs, and provides repair criteria and inspection requirements applicable to the RSGs, does not adversely affect the SG tube design or operating environment. SG tube integrity will continue to be maintained by implementing the TS 5.5.2.11 SG Program to manage SG tube inspection, assessment, and plugging. The requirements established by the TS 5.5.2.11 SG Program are consistent with those in the applicable design codes and standards.

For the change to the peak containment post-accident pressure, the proposed change increases the calculated peak containment internal pressure for the LOCA events from 45.9 psig to 48.0 psig. The revised post-LOCA peak containment pressure is bounded by the existing and revised post-MSLB peak containment pressure. The post-MSLB peak containment pressure decreases from 56.5 psig to 51.5 psig. The proposed peak containment internal pressure for the MSLB accident is less than the containment design pressure of 60 psig and less than the previously calculated pressure.

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

Based on the above, SCE concludes that the proposed amendments present 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: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: June 27, 2008.

Description of amendment request: SONGS Units 2 and 3 requests adoption of an approved change to the standard technical specifications (STS) for Combustion Engineering Pressurized Water Reactor (PWR) Plants (NUREG-1432) and plant-specific technical specifications (TS), to allow replacing the departure from nucleate boiling (DNB) parameter limits with references to the core operating limits report (COLR) in accordance with Generic Letter 88-16, “Removal of Cycle Specific Parameter Limits from Technical Specifications,” dated October 4, 1988. The changes are consistent with NRC approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-487, Revision 1, using the consolidated line-item improvement process (CLIIP).

The NRC staff issued a notice of availability in the Federal Register on June 5, 2007 (72 FR 31108), including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the CLIIP process. The licensee affirmed the applicability of the model NSHC determination in its application dated June 27, 2008.

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: Does the Proposed Change Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated?

Response: No.

The proposed amendment replaces the limit values of the reactor coolant system (RCS) DNB parameters (i.e., pressurizer pressure, RCS cold leg temperature, and RCS flow rate) in TS with references to the COLR, in accordance with the guidance of Generic Letter 88-16, to allow these parameter limit values to be recalculated without a license amendment. The proposed amendment does not involve operation of any required structures, systems, or components (SSCs) in a manner or configuration different from those previously recognized or evaluated. The cycle-specific values in the COLR must be calculated using the NRC-approved methodologies listed in TS 5.6.3, “Core Operating Limits Report (COLR).” Replacing the RCS DNB parameter limits in TS with references to the COLR will maintain existing operating fuel cycle analysis requirements. Because these parameter limits are determined using the NRC approved methodologies, the acceptance criteria established for the safety analyses of various transients and accidents will continue to be met. Therefore, neither the probability nor consequences of any accident previously evaluated will be increased by the proposed change.

Therefore, operation of the facility in accordance with the proposed amendment does not involve a significant increase in the probability or consequences of an accident preciously evaluated.

Criterion 2: Does the Proposed Change Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated?

Response: No.

The proposed amendment to replace the RCS DNB parameter limits in TS with references to the COLR does not involve a physical alteration of the plant, nor a change or addition of a system function. The proposed amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the proposed change. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3: Does the Proposed Change Involve a Significant Reduction in the Margin of Safety?

Response: No.

The proposed amendment to replace the RCS DNB parameter limits in TS with references to the COLR will continue to maintain the margin of safety. The DNB parameter limits specified in the COLR will be determined based on the safety analyses of transients and accidents, performed using the NRC-approved methodologies that show that, with appropriate measurement uncertainties of these parameters accounted for, the acceptance criteria for each of the analyzed transients are met. This provides the same margin of safety as the limit values currently specified in the TS. Any future revisions to the safety analyses that require prior NRC approval are identified per the 10 CFR 50.59 review process.

Therefore, the proposed amendment would not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

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

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

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

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

Date of application for amendments: October 26, 2007.

Brief description of amendments: The proposed amendment would revise the Technical Specification requirements related to control room envelope habitability in accordance with the NRC-approved Revision 3 of Technical Specification Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-448, “Control Room Habitability.”

Date of publication of individual notice in the Federal Register : August 29, 2008 (73 FR 51014).

Expiration date of individual notice: September 29, 2008 (Public Comments) and October 28, 2008 (Requests for Hearing).

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 theseamendments 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 51.22(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 Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: September 27, 2007.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) TS 3.7.2, “Main Steam Isolation Valves,” and TS 3.7.3, “Main Feedwater Isolation Valves, Main Feedwater Control Valves, Associated Bypass Valves and Tempering Valves,” by removing the specific isolation time for the isolation valves from the associated surveillance requirements.

Date of issuance: September 8, 2008.

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

Amendment Nos.:244 and 238.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register : February 26, 2008 (73 FR 10 10297).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 8, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 13, 2008.

Brief description of amendment: The amendment replaces the current Arkansas Nuclear One, Unit No. 2 (ANO-2) TS 3.4.8, “RCS [reactor coolant system] Specific Activity,” limit on RCS gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent Xe-133 (DEX) definition that would replace the current E Bar average disintegration energy definition. In addition, the current dose equivalent I-131 (DEI) definition would be revised to allow the use of additional thyroid dose conversion factors (DCFs). This request adopted Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, “Deletion of E Bar Definition and Revision to RCS [reactor coolant system] Specific Activity Technical Specification'' (Agencywide Documents Access and Management System Accession No. ML052630462), for pressurized water reactor Standard Technical Specifications (STS) for ANO-2.

Date of issuance: September 8, 2008.

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

Amendment No.: Unit 2-282.

Renewed Facility Operating License No. NPF-6: Amendment revised the Technical Specifications and license.

Date of initial notice in Federal Register : May 6, 2008 (73 FR 25039).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 8, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 22, 2007, as supplemented by letters dated April 22, and July 8, 2008.

Brief description of amendment: The amendment revises Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.0.4 and Surveillance Requirement (SR) 4.0.4 to adopt the provisions of Industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints.” This operating license improvement was made available by the U.S. Nuclear Regulatory Commission (NRC) on April 4, 2003, as part of the consolidated line item improvement process. The proposed TS changes also include an additional application of LCO 3.0.4.c for TS 3.4.3, “Pressurizer Spray Valves.”

Date of issuance: August 28, 2008.

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

Amendment No.: Unit 2-281.

Renewed Facility Operating License No. NPF-6: Amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register : December 18, 2007 (72 FR 71710). The supplements dated April 22, and July 8, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 28, 2008.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois

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

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

Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353, Limerick Generating Station, Unit 1 and 2, Montgomery County, Pennsylvania

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

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

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

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

Date of application for amendments: August 8, 2007.

Brief description of amendments: The amendment replaces references to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) in the applicable technical specification (TS) section for the Inservice Testing Program (IST) for the Exelon Generation Company, LLC, and AmerGen Energy Company, LLC, plants that have implemented industry Improved Technical Specifications. The changes are based on Technical Specification Task Force (TSTF) 479, Revision 0, “Changes to Reflect Revision of 10 CFR 50.55a.” For all units except Oyster Creek and TMI-1, the amendments also incorporate TSTF-497, Revision 0, “Limit Inservice Testing Program SR [Surveillance Requirement] 3.0.2 Application to Frequencies of 2 Years or Less,” which adds a provision in the applicable TS section to only apply the extension allowance of SR 3.0.2 to the frequency table listed in the TS as part of the IST program and to normal and accelerated inservice testing frequencies of two years or less, as applicable.

Date of issuance: August 28, 2008.

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

Amendment Nos.:153, 153, 157, 157, 229, 222, 194, 155, 268, 268, 272, 241, 236 and 266.

Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66, DPR-19, DPR-25, NPF-39, NPF-85, DPR-16, DPR-44, DPR-56, DPR-29, DPR-30, and DPR-50: The amendments revised the Technical Specifications/Licenses.

Date of initial notice in Federal Register : December 4, 2007 (72 FR 68213). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 28, 2008.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendment: August 24, 2007, supplemented by letter dated June 11, 2008.

Brief description of amendment: The amendments consist of changes to the technical specifications of each unit, increasing the minimum required volume of fuel oil in the emergency diesel generator day tanks from 200 gallons to 250 gallons.

Date of issuance: August 27, 2008.

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

Amendment Nos.:193 and 154.

Facility Operating License Nos. NPF-39 and NPF-85. These amendments revised the license and the technical specifications.

Date of initial notice in Federal Register : June 20, 2008 (73 FR 35168). The NRC staff's original proposed no significant hazards determination was based on the supplement dated June 11, 2008. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 27, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 13, 2007, as supplemented on February 28, 2008, March 28, 2008, April 17, 2008, May 23, 2008, July 29, 2008, August 7, 2008, andAugust 21, 2008.

Brief description of amendments: The amendments modify the Technical Specifications to support application of Alternative Source Term (AST) methodology at PBAPS Units 2 and 3. The fission product release from the reactor core into containment is referred to as the “source term,” and is characterized by the composition and magnitude of the radioactive material, the chemical and physical properties of the material, and the timing of the release from the reactor core as discussed in Technical Information Document (TID) 14844, “Calculation of Distance Factors for Power and Test Reactor Sites.” Since the publication of TID 14844, advances have been made in understanding the composition and magnitude, chemical form, and timing of fission product releases from severe nuclear power plant accidents. In light of these insights, NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants,” was published in 1995 with revised ASTs for use in the licensing of future light-water reactors.

The Nuclear Regulatory Commission (NRC), in Title 10 of the Code of Federal Regulations, Section 50.67 (10 CFR 50.67), “Accident source term,” subsequently allowed the use of the ASTs described in NUREG-1465 at operating plants. This request to apply the AST methodology is made in accordance with 10 CFR 50.67, with the exception that TID 14844 will continue to be used as the radiation dose basis for equipment qualification at PBAPS Units 2 and 3. Application of the AST methodology at PBAPS Units 2 and 3 requires that radiation dose limits specified in 10 CFR 50.67 are adhered to for the exclusion area boundary, the low population zone outer boundary, and the facility control room.

Date of issuance: September 5, 2008.

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

Amendment Nos.:269 and 273.

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

Date of initial notice in Federal Register : May 6, 2008 (73 FR 25040). The supplements dated February 28, 2008, March 28, 2008, April 17, 2008, May 23, 2008, July 29, 2008, August 7, 2008, and August 21, 2008, clarified the application, did not expand the scope of the application as originally noticed, and did not change the initial proposedno significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 5, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: February 20, 2008.

Brief description of amendment: This amendment revised an Applicability footnote in Technical Specification (TS) Table 3.3.2.1-1, “Control Rod Block Instrumentation,” to permit use of an improved optional Banked Position Withdrawal Sequence (BPWS) reactor shutdown process. Corresponding changes are in accordance with the Bases of TS 3.1.6, “Control Rod Pattern,” and the Bases of TS 3.3.2.1, to reference the new BPWS shutdown method. This amendment is consistent with Technical Specification Task Force (TSTF) Traveler TSTF-476-A, Revision 1, “Improved BPWS Control Rod Insertion Process (NEDO-33091),” and the Consolidated Line Item Improvement Process Notice of Availability dated May 23, 2007.

Date of issuance: August 28, 2008.

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

Amendment No.:150.

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

Date of initial notice in Federal Register : April 22, 2008 (73 FR 21659).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 28, 2008.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1 (NMP1), Oswego County, New York

Date of application for amendment: September 27, 2007, as supplemented by letter dated June 5, 2008.

Brief description of amendment: The amendment changes the NMP1 Technical Specifications (TSs) by revising the operability requirements contained in TS Section 3.2.7, “Reactor Coolant System Isolation Valves,” and associated requirements contained in TS Section 3.6.2, “Protective Instrumentation.” The amendment will modify the conditions for which reactor coolant system isolation valves (RCSIVs) and associated isolation instrumentation must be operable to include the hot shutdown reactor operating condition. In addition, it will be required that the RCSIVs in the shutdown cooling (SDC) system and associated isolation instrumentation be operable during the cold shutdown reactor operating condition and the refueling reactor operating condition. Lastly, TS Section 3.6.2 (Table 3.6.2b) will be revised to delete unnecessary operability requirements for the cleanup system and SDC system high area temperature isolation instrumentation, consistent with the proposed revisions to the RCSIV operability requirements.

Date of issuance: August 27, 2008.

Effective date: As of the date of issuance to be implemented within 90 days.

Amendment No.:197.

Renewed Facility Operating License No. DPR-63: Amendment revised the License and TSs.

Date of initial notice in Federal Register : November 20, 2007 (72 FR 65367). The supplement dated June 5, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 27, 2008.

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 amendment request: November 5, 2007, as supplemented April 7, 2008.

Brief description of amendment request: TS Section 5.5.17, “Containment Leakage Rate Testing Program,” is changed to resolve a timing conflict between the FNP, Unit 2 R20 refueling outage schedule and the 15-year test date for the FNP, Unit 2 Type A Containment Integrated Leak Rate Test (ILRT). Although Unit 1 does not have a current timing conflict, a similar Unit 1 change was requested for consistency. The change adds approximately 1 month to the previously approved required date.

Date of issuance: September 2, 2008.

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

Amendment Nos.: Unit 1-177; Unit 2-170.

Facility Operating License Nos. NPF-2 and NPF-8: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : January 29, 2008 (73 FR 5229).

The supplement dated April 7, 2008, provided clarifying information that did not change the scope of the application or the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 2, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 29, 2006, as supplemented November 6, November 27, 2006, January 30, June 22, July 16, August 13, October 18, December 11, 2007, January 24, February 4, February 25 (two letters, nos. 1389 and 0175), February 27, March 13, April 1, May 5, June 25, July 2, July 14, and August 14, 2008.

Brief description of amendments: The amendments revise the licensing basis with a full scope implementation of an alternative source term (AST) for HNP.

Date of issuance: August 28, 2008.

Effective date: As of the date of issuance and shall be implemented by May 31, 2012 for Hatch Unit 1 and by May 31, 2011, for Hatch Unit 2.

Amendment Nos.: Unit 1-256, Unit 2-200.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register : July 23, 2008 (73 FR 42834).

The supplement dated August 14, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 28, 2008.

No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. Within 60 days after the date of publication of this notice, person(s) 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 via electronic submission through the NRC E-Filing system 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 person(s) 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 basesfor the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. [1] Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

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

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

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 hearing or a petition for leave to intervene must be filed in accordance with the NRC E-Filing rule, which the NRC promulgated in August 28, 2007, (72 FR 49139). The E-Filing process requires participants to submit and serve documents over the Internet or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek a waiver in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at HEARINGDOCKET@NRC.GOV, or by calling (301) 415-1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/ requestor will need to download the Workplace Forms Viewer [TM] to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms Viewer [TM] is free and is available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.

Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The EIE system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically may seek assistance through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line, which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, Monday through Friday. The help line number is (800) 397-4209 or locally, (301) 415-4737.

Participants who believe that they have a good cause for not submitting documents electronically must file a motion, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted 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; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service.

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later than11:59 p.m. Eastern Time on the due date.

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a Presiding Officer. Participants are requested not to include personal privacy information, such as Social Security numbers, home addresses, or home phone numbers in their filings. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: August 26, 2008, as supplemented on August 28, 2008.

Description of amendment request: The amendments revise Functional Unit 6.f of Table 3.3-3, “Engineered Safety Feature Actuation System Instrumentation,” modifying the mode of applicability with two footnotes. The first footnote indicates that the auxiliary feedwater (AFW) auto-start function associated with the trip of main feedwater (MFW) pumps in Mode 2 is only required when one or more MFW pumps are supplying feedwater to the steam generators. The second footnote, which annotates the minimum channels operable column for Functional Unit 6.f of TS Table 3.3-3, indicates that one channel may be inoperable during Mode 1 for up to 4 hours when starting up or shutting down a MFW pump. Functional Unit 6.f of technical specification Table 3.3-3 is an anticipatory trip function that provides early actuation of the AFW system.

Date of issuance: August 29, 2008.

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

Amendment Nos:319 and 312.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the technical specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): No. The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated August 29, 2008.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: Thomas H. Boyce.

Dated at Rockville, Maryland, this 11th day of September 2008.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

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.

References

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