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Notice

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

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Information about this document as published in the Federal Register.

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

Start Preamble

I. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding 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 14, 2008, to August 27, 2008. The last Start Printed Page 52413biweekly notice was published on August 26, 2008 (73 FR 50356).

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 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, 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's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

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

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

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.Start Printed Page 52414

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 ViewerTM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms ViewerTM 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 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.

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.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power Station (KPS), Kewaunee County, Wisconsin

Date of amendment request: August 14, 2008.

Description of amendment request: The proposed amendment would modify Specification 4.4.f.1, “Containment Isolation Device Verification,” of the Technical Specifications (TS) to require verification that the 36-inch containment purge and vent isolation valves are sealed closed when the reactor is at greater than Cold Shutdown conditions.

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 Design Bases Accidents (DBA) that result in a release of radioactive material within containment are a steam line break, rupture of a rod cluster control assembly, and loss-of-coolant accident (LOCA). In the analyses for each of these accidents, it is assumed that containment isolation valves are either closed or function to close within the required isolation time following accident initiation. This ensures that potential leakage paths to the environment Start Printed Page 52415through containment isolation valves (including containment purge and vent isolation valves) are minimized. The safety analyses assume that the containment purge and vent isolation valves are closed at accident initiation.

The safety function of the containment purge and vent isolation valves is to support the Containment Isolation system by confining fission products within the Primary Containment system boundary during a DBA. The proposed amendment would require verification that the containment purge and vent isolation valves are sealed closed when the reactor is at greater than Cold Shutdown conditions. This requirement ensures the valves are in their required DBA post-accident position when the reactor is at greater than Cold Shutdown conditions.

Verifying the containment purge and vent isolation valves are sealed closed at 31-day intervals does not add, delete, or modify any KPS system, structure, or component (SSC). Verifying that the containment purge and vent isolation valves are sealed closed when the reactor is at greater than Cold Shutdown conditions has no adverse effect on the ability of the plant to mitigate the effects of DBAs. The subject surveillance requirement constitutes a verification of isolation valve position and has no effect on equipment. Verification of valve closure only ensures the previous assumptions made in evaluating the consequences of DBAs remain valid. Therefore, there is no increase in the probability of an accident by performing the surveillance in additional modes of plant operation.

Therefore, the proposed amendment does 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.

Verifying the containment purge and vent isolation valves are sealed closed when the reactor is at greater than Cold Shutdown conditions at 31-day intervals ensures these valves are in their required DBA post-accident position when the design function is required. The proposed amendment does not change the manner in which these valves are operated when the reactor is at or below Cold Shutdown or their design function. The proposed amendment does not create any new failure mechanisms or malfunctions for plant equipment or the nuclear fuel.

In addition, the containment purge and vent isolation valves are not accident initiators. Their function is only for mitigation of accidents.

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

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

Response: No.

Verifying the containment purge and vent isolation valves are sealed closed when the reactor is at greater than Cold Shutdown conditions at 31-day intervals ensures these valves are in their required DBA post-accident position when the design function is required. The proposed amendment does not change the manner in which these valves are operated when the reactor is at or below Cold Shutdown condition.

The proposed amendment would align the KPS TS with applicable NRC requirements stated in NUREG-0800 [“Standard Review Plan,”], Section 6.2.4 and NUREG-0737 [“Clarification of Three Mile Island Action Plan Requirements,”], Item II.E.4.2. The proposed amendment does not result in altering or exceeding a design basis or safety limit for the plant. The safety analysis of record, including evaluations of the radiological consequences of design basis accidents, will remain applicable and unchanged.

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

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

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.

NRC Branch Chief: Lois James.

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: August 1, 2008.

Description of amendment request: The proposed amendments would authorize changes to the Updated Final Safety Analysis Report (UFSAR) to account for small areas of carbon steel (CS) and low alloy steel that may be exposed to the reactor coolant system (RCS).

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?

No.

The Pressurizer vent nozzle and thermowell, as components of the RCS, must maintain system pressure boundary. RCS design pressure is 2500 psig and design temperature is 670 °F. The vent nozzle and thermowell replacements are designed for the RCS pressure and temperature. As described above, the material of the new Pressurizer vent nozzle and thermowell is an improvement in the PWSCC [primary water stress corrosion cracking] resistance of those components as compared to the original components. The design of the new Pressurizer vent nozzle and thermowell exposes small areas of the Pressurizer shell carbon steel to a stagnant reactor coolant environment. However, the corrosion of the Pressurizer shell is considered negligible. Therefore, the replacement of the Pressurizer vent nozzle and thermowell do not more than minimally increase the likelihood of occurrence of a malfunction. Corrosion evaluations performed show that all applicable ASME Code requirements are met.

It is concluded that the consequences of a Pressurizer vent nozzle or Pressurizer thermowell failure resulting in a LOCA [loss-of-coolant accident] are bounded by existing analysis. Therefore, there is no increase in the probability or consequences of an accident.

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

No.

The only credible accident involving the failure of these components is bounded by existing LOCA analyses. There are no new accidents that need to be postulated due to the replacement of the Pressurizer vent nozzle and Pressurizer thermowell. Therefore, this proposed activity will not create the possibility of a new or different kind of accident from any kind of accident previously evaluated.

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

No.

The mitigation technique selected for the Pressurizer vent nozzle and the Pressurizer thermowell exposes a small area of CS to the RCS environment. As required by the ASME Code, Section III, a supporting corrosion evaluation was developed within each of the two component designs. The technical package for the replacement of the Pressurizer vent nozzle and the Pressurizer thermowell utilized calculations to support the evaluation of the acceptability of this repair/replacement activity. The corrosion evaluation for the Pressurizer vent resulted in a conservative general stagnant corrosion rate of 0.0018 inches per year and the corrosion evaluation for the Pressurizer thermowell resulted in a conservative general corrosion rate of 0.00142 inches per year. The critical corrosion distance is the radius from the exposed CS surface to the edge of the weld pad. This distance is at least 1.1 inches for both the vent and thermowell designs. With this distance, a corrosion rate of less than 2 mils per year is not significant when compared to the 60 year component design life, which begins at the time of installation.

The original Pressurizer was designed to meet Section III of the ASME Code, and the Pressurizer, as modified, meets Section III of the ASME code. Although this change does expose small areas of CS in the Pressurizer, 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 Start Printed Page 52416review, 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.

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

Date of amendment request: July 8, 2008.

Description of amendment request: The proposed amendment to Indian Point Nuclear Generating Units Nos. 2 and 3 (IP2 and IP3) would require the licensee to submit information and analyses associated with extending the Reactor Vessel (RV) Inservice Inspection (ISI) Interval from 10 to 20 years for specific pressure retaining welds in the RV.

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 will revise the license to require the submission of information and analyses to the NRC following completion of each ASME [American Society of Mechanical Engineers] [C]ode, Section XI, Category B-A and B-D Reactor Vessel weld inspection. The extension of the ISI from 10 to 20 years is being evaluated as part of the relief request independent from the license change. Submission of the information and analyses can have no effect on the consequences of an accident or the probability of an accident because the submission of information is not related to the operation of the plant or any equipment, the programs and procedures used to operate the plant, or the evaluation of accidents. The submittal of information and analyses provides the opportunity for the NRC to independently assess the information and analyses.

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 will only affect the requirement to submit information and analyses when specified inspections are performed. There are no changes to plant equipment, operating characteristics or conditions, programs, and procedures or training. Therefore, there are no potential new system interactions or failures that could 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 will revise the license to require the submission of information and analyses to the NRC following completion of each ASME [C]ode, Section XI, Category B-A and B-D Reactor Vessel weld inspection which does not affect any Limiting Conditions for Operation used to establish the margin of safety. The requirement to submit information and analyses is an administrative tool to assure the NRC has the ability to independently review information developed by the [l]icensee. 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. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Mark G. Kowal.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: July 9, 2008.

Description of amendment request: The proposed amendment will revise the test acceptance criteria specified in the Technical Specification Surveillance Requirement (SR) 3.8.1.10 for the Diesel Generator (DG) endurance test. The load ranges and power factors specified for the test will be changed for consistency with the associated safety analyses.

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?

No. The proposed change revises the acceptance criteria to be applied to an existing surveillance test of the facility emergency diesel generators (DGs). Performing a surveillance test is not an accident initiator and does not increase the probability of an accident occurring. The proposed new acceptance criteria will assure that the DGs are capable of carrying the peak electrical loading assumed in the various existing safety analyses which take credit for the operation of the DGs. Establishing acceptance criteria that bound existing analyses validates the related assumption used in those analyses regarding the capability of equipment to mitigate accident conditions. 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?

No. The proposed change revises the test acceptance criteria for a specific performance test conducted on the existing DGs. The proposed change does not involve installation of new equipment or modification of existing equipment, so no new equipment failure modes are introduced. The proposed revision to the DG surveillance test acceptance criteria also is not a change to the way that the equipment or facility is operated and no new accident initiators are created. 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?

No. The conduct of performance tests on safety-related plant equipment is a means of assuring that the equipment is capable of maintaining the margin of safety established in the safety analyses for the facility. The proposed change in the DG technical specification surveillance test acceptance criteria is consistent with values assumed in existing safety analyses is consistent with the design rating of the DGs. Therefore the propose 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. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Mark G. Kowal.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of amendment request: May 5, 2008.

Description of amendment request: The proposed amendment would correct an error in Section A.1 of the renewed operating license and remove several outdated license conditions relating to surveillance requirements. Specifically, it would remove the words “filed by Entergy Nuclear Palisades, LLC (ENP) and Entergy Nuclear Start Printed Page 52417Operations, Inc. (ENO)” in Section A.1, spell-out acronyms used in Section 1.F, and delete license conditions 2.C.(4) and 2.C.(5), and delete Table 2.C.(5).

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 license amendment deletes incorrect or outdated information from the renewed facility operating license. The proposed amendment does not involve operation of the required structures, systems or components (SSCs) in a manner or configuration different from those previously recognized or evaluated.

Modification of renewed facility operating license sections 1.A and 1.F and deletion of license conditions 2.C.(4), 2.C.(5), and Table 2.C.(5) is administrative and has no impact on plant operation or equipment.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously 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 license amendment does not involve a physical alteration of any SSC or change the way any SSC is operated. The proposed license amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated.

Modification of renewed facility operating license sections 1.A and 1.17 and deletion of license conditions 2.C.(4), 2.C.(5), and Table 2.C.(5) is administrative and has no impact on plant operation or equipment.

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.

Modification of renewed facility operating license sections 1.A and 1.F and deletion of license conditions 2.C.(4), 2.C.(5), and Table 2.C.(5) is administrative and has no impact on plant operation or equipment or on any margin of safety.

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 request involves no significant hazards consideration.

Attorney for licensee: Mr. William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: Lois M. James.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of amendment request: May 5, 2008

Description of amendment request: The proposed amendment would revise renewed facility operating license DPR-20 to remove license condition 2F. This license condition describes reporting requirements for exceeding the facility steady-state reactor core power level described in condition 2.C.(1). The proposed change is consistent with the Nuclear Regulatory Commission (NRC)-approved change notice published in the Federal Register on November 4, 2005, announcing the availability of this improvement through the consolidated line item improvement process. The Federal Register Notice included a model safety evaluation and model no significant hazards consideration (NSHC) determination, relating to the elimination of the license condition involving reporting of violations of other requirements (typically in License Conditions 2.C) in the operating license of some commercial nuclear power plants. The licensee affirmed the applicability of the model NSHC determination in its application dated May 5, 2008.

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

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

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 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. William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: Lois M. James.

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

Date of amendment request: July 29, 2008.

Description of amendment request: The proposed amendments would remove time, cycle, or modification-related items from the operating licenses (OLs) and technical specifications (TSs) at both stations. Additionally, the proposed amendments would correct typographical errors introduced into the TSs at both stations in previous amendments. The time, cycle, or modification-related items have been implemented or superseded, are no longer applicable, and no longer need to be maintained in their associated OLs or TSs.

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 initial conditions and methodologies used in the accident analyses remain unchanged. The proposed changes do not change or alter the design assumptions for the systems or components used to mitigate the consequences of an accident. Therefore, accident analyses results are not impacted.

All changes proposed by EGC in this amendment request are administrative in nature, and are removing one-time requirements that have been satisfied or items that are no longer applicable. There are no physical changes to the facilities, nor any changes to the station operating procedures, Start Printed Page 52418limiting conditions for operation, or limiting safety system settings.

Based on the above discussion, 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.

None of the proposed changes affect the design or operation of any system, structure, or component in the plant. The safety functions of the related structures, systems, or components are not changed in any manner, nor is the reliability of any structure, system, or component reduced by the revised surveillance or testing requirements. The changes do not affect the manner by which the facility is operated and do not change any facility design feature, structure, system, or component. No new or different type of equipment will be installed. Since there is no change to the facility or operating procedures, and the safety functions and reliability of structures, systems, or components are not affected, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Based on this evaluation, the proposed changes do 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 to the Facility Operating Licenses and TS are administrative in nature and have no impact on the margin of safety of any of the TS. There is no impact on safety limits or limiting safety system settings. The changes do not affect any plant safety parameters or setpoints. The Operating License Conditions have been satisfied as required. There are no changes to the conditions themselves.

Based on this evaluation, the proposed change does not involve a significant reduction in a margin of safety.

The Nuclear Regulatory Commission (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. Bradley J. Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell Gibbs.

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

Date of amendment request: January 23, 2008.

Description of amendment request: Replace the current Technical Specification pressure/temperature (P/T) limit curves with new P/T limit curves applicable to 55 effective full-power years (EFPY). The low-temperature overpressure protection (LTOP) requirements, which are based on the P/T limits, will also be applicable to 55 EFPY.

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) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes have been determined in accordance with the methodologies set forth in the regulations to provide an adequate margin of safety to ensure that the reactor vessel will withstand the effects of normal startup and shutdown cyclic loads due to system temperature and pressure changes as well as the loads associated with reactor trips. The regulations of 10 CFR Part 50 Appendix A, Design Criterion 14 and Design Criterion 31 remains satisfied. The pressure-temperature (P/T) limit curves in the Technical Specifications are conservatively generated in accordance with the fracture toughness requirements of the ASME [American Society of Mechanical Engineers] Code Section XI, Appendix G. The margins of safety against fracture provided by the P/T limits using the requirements of 10 CFR 50 Appendix G are equivalent to those recommended in ASME Section XI, Appendix G. The Adjusted Reference Temperature (ART) values are based on the guidance of RG [Regulatory Guide] 1.99 [Reference 4].

The proposed changes will not result in physical changes to structures, systems or components SSCs or to event initiators or precursors. Changing the heatup and cooldown curves and the pressure relief setpoints to reflect 55 EFPY does not affect the ability to control the RCS [reactor coolant system] at low temperatures such that the integrity of the reactor coolant pressure boundary would not be compromised by violating the P/T limits.

The proposed changes will not impact assumptions and conditions previously used in the radiological consequence evaluations nor affect mitigation of these consequences due to an accident described in the UFSAR [Updated Final Safety Analysis Report]. Also, the proposed changes will not impact a plant system such that previously analyzed SSCs might be more likely to fail. The initiating conditions and assumptions for accidents described in the UFSAR remain as analyzed.

Thus, based on the above, reasonable assurance is provided that the proposed amendment does not significantly increase the probability or consequences of accidents previously evaluated.

(2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The requirements for P/T limit curves and LTOP have been in place since the beginning of plant operation. The revised curves are based on a later edition of Section XI of the ASME Code that incorporates current industry standards for P/T curves. The revised curves also are based on reactor vessel irradiation damage predictions using RG 1.99 methodology. No new failure modes are identified nor are any SSCs required to be operated outside of their design bases. Consequently, the proposed amendment 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.

The proposed P/T curves continue to maintain the safety margins of 10 CFR 50 Appendix G by defining the limits of operation which prevent nonductile failure of the reactor pressure vessel. Analyses have demonstrated that the fracture toughness requirements are satisfied and that conservative operating restrictions are maintained for the purpose of low temperature overpressure protection. The P/T limit curves provide assurance that the RCS pressure boundary will behave in a ductile manner and that the probability of a rapidly propagating fracture is minimized. Therefore, operation in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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: Thomas H. Boyce.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

Date of amendment request: April 4, 2008.

Description of amendment request: The licensee proposed to change the Technical Specifications (TS) to revise requirements for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9. This LCO would establish conditions under which systems would remain operable when required physical barriers are not capable of providing their related support function. This proposed amendment is consistent with the NRC's approved Technical Specification Task Start Printed Page 52419Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-427, Revision 2. A notice of availability of this TS improvement was published in the Federal Register on October 3, 2006 (71 FR 58444) as part of NRC's Consolidated Line Item Improvement Process (CLIIP).

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee provided an analysis of the issue of no significant hazards consideration by citing the proposed NSHC determination published by the NRC staff in the Federal Register referenced above. That proposed NSHC is reproduced below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the 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 the allowance provided by proposed LCO 3.0.9 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 3.0.9. 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 an unavailable barrier, if risk is assessed and managed, 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 unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the 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. This application of LCO 3.0.9 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP [incremental conditional core damage probability] and ICLERP [incremental conditional large early release probability] ) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the analysis cited by the licensee, and has found that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the proposed amendment involves 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 Acting Branch Chief: Lois M. James.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

Date of amendment request: April 22, 2008.

Description of amendment request: The licensee proposed to change the Technical Specifications (TS) to (1) revise the surveillance requirement frequency in Specification 3.1.3, “Control Rod Operability,” to require control rod notch testing to be performed at a 31-day frequency for both partially and fully withdrawn control rods; and (2) revise Example 1.4-3 in Section 1.4, “Frequency,” to clarify the applicability of the 1.25 surveillance test interval extension. These proposed changes are consistent with the NRC's approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications (STS) Change Traveler, TSTF-475, Revision 1. A notice of availability of this TS improvement was published in the Federal Register on November 13, 2007 (72 FR 63935), as part of the NRC's Consolidated Line Item Improvement Process (CLIIP).

Basis for proposed no significant hazards consideration determination: As required by 10 FR 50.91(a), the licensee provided an analysis of the issue of no significant hazards consideration by citing the proposed NSHC determination published by the NRC staff in the Federal Register notice referenced above. That proposed NSHC is reproduced below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change generically implements TSTF-475, Revision 1, “Control Rod Notch Testing Frequency and SRM [Source Range Monitoring] Insert Control Rod Action.” TSTF-475, Revision 1, modifies NUREG-1433 (BWR [Boiling Water Reactor]/4) and NUREG-1434 (BWR/6) STS. The changes (1) revise TS testing frequency for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, “Control Rod OPERABILITY,”, and (2) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension. The consequences of an accident after adopting TSTF-475, Revision 1 are no different than the consequences of an accident prior to adoption. 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 Accident Previously Evaluated

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change 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 analyzed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2 frequency in TS 3.1.3, “Control Rod OPERABILITY,” and (2) revise Example 1.4-3 in Section 1.4 “Frequency” to clarify the applicability of the 1.25 surveillance test interval extension. The GE Nuclear Energy Report, “CRD Notching Surveillance Testing for Limerick Generating Station,” dated November 2006, concludes that extending the control rod notch test interval from weekly to monthly is not expected to impact the reliability of the scram system and that the analysis supports the decision to change the surveillance frequency. Therefore, the proposed changes in TSTF-475, Revision 1 Start Printed Page 52420are acceptable and do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the analysis cited by the licensee, and has found that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the proposed amendment involves 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 Acting Branch Chief: Lois M. James.

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.

Description of amendment request: The proposed amendments would amend the Facility Operating Licenses by revising the licensing basis loss of coolant accident and main steam line break accident radiological dose consequences for Prairie Island Nuclear Generating Plant, Units 1 and 2, as currently described in the Updated Safety Analysis Report Section 14.5 and Section 14.9. This proposed amendment also proposes concomitant amendments to Appendix A of the Facility Operating Licenses, Technical Specifications (TS) 3.3.5, “Containment Ventilation Isolation Instrumentation”, 3.4.17, “RCS [Reactor Coolant System] Specific Activity”, and 3.6.3, “Containment Isolation Valves”, which are necessary to implement the proposed revised analyses.

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.

This license amendment request proposes implementing revised loss of coolant accident and main steam line break accident dose consequence analyses to address modeling nonconservatisms and update the analyses for new fuel types and provide margin for power uncertainty. These analyses assumed that the containment inservice purge system penetrations are isolated, thus this license amendment request proposes Technical Specification revisions which will require these penetrations to be blind flanged during plant operations; these changes allow the Technical Specification requirements for containment ventilation isolation instrumentation to be removed. This license amendment request also proposes associated more restrictive limits in the Technical Specification for reactor coolant system specific activity since the main steam line break accident analysis assumed lower limits.

The accident radiological dose consequences analyses inputs, methodologies and outputs modified by this request are not accident initiators and do not affect the frequency of occurrence of previously analyzed transients. Likewise, the reactor coolant system specific activity limits are not accident initiators and do not affect the frequency of occurrence of previously analyzed transients.

The containment inservice purge system is not an accident initiator and therefore removal of its Technical Specifications does not involve an increase in the probability of an accident. The Technical Specification changes proposed in this license amendment request require the containment inservice purge system to be blind flanged during Modes 1, 2, 3, and 4, therefore removal of the containment ventilation isolation instrumentation Technical Specifications and other Technical Specification system operating requirements does not involve an increase in the consequences of an accident previously evaluated.

The loss of coolant accident and main steam line break accident radiological dose consequences analyses demonstrated the results are within the applicable regulatory limits and guidance using revised inputs, including the proposed lower Technical Specification reactor coolant system specific activity limits, and methodologies. Thus these changes do not involve a significant increase in the consequences of an 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 amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

This license amendment request proposes implementing revised loss of coolant accident and main steam line break accident dose consequence analyses to address modeling nonconservatisms and update the analyses for new fuel types and provide margin for power uncertainty. These analyses assumed that the containment inservice purge system penetrations are isolated, thus this license amendment request proposes Technical Specification revisions which will require these penetrations to be blind flanged during plant operations; these changes allow the Technical Specification requirements for containment ventilation isolation instrumentation to be removed. This license amendment request also proposes associated more restrictive limits in the Technical Specification for reactor coolant system specific activity since the main steam line break accident analysis assumed lower limits.

This license amendment request does not involve physical changes to the plant structures, systems or components and there is no adverse impact on component or system interactions due to the proposed changes. The modes of operation of the plant remain unchanged and the design functions of the safety systems remain in compliance with the applicable safety analysis acceptance criteria. These changes do not create new failure modes or mechanisms and no new accident precursors are generated.

When the containment inservice purge system is not being operated, current Technical Specifications require the system's penetrations to be blind flanged in Modes 1, 2, 3, and 4 to provide post-accident containment integrity. This license amendment proposes to require the system penetrations to be blind flanged at all times during these Modes and prevent operation of the system in these Modes. Since containment integrity is provided with the penetrations blind flanged and this change only extends the time during which the system is in this configuration, these changes do not create the possibility of a new or different kind of accident.

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

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

Response: No.

This license amendment request proposes implementing revised loss of coolant accident and main steam line break accident dose consequence analyses to address modeling nonconservatisms and update the analyses for new fuel types and provide margin for power uncertainty. These analyses assumed that the containment inservice purge system penetrations are isolated, thus this license amendment request proposes Technical Specification revisions which will require these penetrations to be blind flanged during plant operations; these changes allow the Technical Specification requirements for containment ventilation isolation instrumentation to be removed. This license amendment request also proposes associated more restrictive limits in the Technical Specification for reactor coolant system specific activity since the main steam line break accident analysis assumed lower limits.

The loss of coolant accident and main steam line break accident radiological dose consequences analyses have incorporated revised inputs, including the proposed lower Technical Specification reactor coolant system specific activity limits, and utilized revised methodologies. The results of these revised analyses satisfy the applicable regulatory limits and guidance. There is no adverse effect on plant safety due to this proposed license amendment.

The containment inservice purge system is not credited for mitigation of any accidents or any other safety function, thus, removal of its associated Technical Specifications does not involve reduction in a margin of safety. The containment ventilation isolation instrumentation system is credited for isolation of the containment inservice purge system following an accident and the valves are assumed to meet containment integrity Start Printed Page 52421leakage rate limits. This license amendment request proposes to require the containment inservice purge system containment penetrations to be blind flanged during Modes 1, 2, 3, and 4 and the blind flanged penetrations will be required to meet containment integrity leakage rate limits. With these changes, containment integrity is maintained in accordance with the current Technical Specification requirements, thus, this change does not involve reduction in a margin of safety.

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 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.

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

Date of amendment request: July 30, 2008.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.8.3, “Onsite Power Distribution Systems,” to establish a separate TS Action statement for inoperable inverters associated with the 120 volt alternating current (VAC) distribution panels. The intent of the proposed amendment is to extend the allowed outage time for inoperable inverters from 8 hours to 24 hours.

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 inverters and associated 120 VAC distribution panels are not initiators to any accident sequence analyzed in the Updated Final Safety Analysis Report (UFSAR).

The proposed change does not increase the number of inverters permitted to be inoperable at one time. With one or both inverters inoperable in a single channel, sufficient capacity and capability remain to assure required safety functions can be performed. The proposed 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 are unaffected by these proposed changes. The likelihood of previously analyzed failures remains unchanged.

Therefore, the proposed change does 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.

No physical changes will be made to the plant or how the plant is operated. As such, no new or different kind of accident due to a credible new failure mechanism, malfunction, or accident initiator will be created as a result of this proposed change. Any alteration in procedures will continue to ensure that the plant remains within analyzed limits, and no change is required to the procedures relied upon to respond to an off-normal event as described in the UFSAR.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change would extend the allowed outage time for one or two inoperable inverters in a single channel. The proposed change does not increase the number of inverters permitted to be inoperable at one time. There is no change to any design basis or safety limits. Operation in accordance with the proposed TS ensures that the 120 VAC instrument distribution system is capable of performing its functions as described in the UFSAR.

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

NRC Branch Chief: Harold K. Chernoff.

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.

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

Date of application for amendment: July 13, 2007, as supplemented by letters dated July 13, September 12, November 19, December 13, and December 17, 2007; January 10 (4 letters), January 11 (4 letters), January 14, and January 18 (5 letters), January 31, February 25 (2 letters), March 5, March 10 (2 letters), March 25, March 27, April 4, April 24, April 29, May 15, May 20, May 21, July 10, and July 16, 2008.

Brief description of amendment: The amendment increased the Millstone Power Station, Unit No. 3 (MPS3) maximum steady-state reactor core power level from the previous licensed thermal power level of 3,411 megawatts thermal (MWt) to 3,650 MWt, which is an increase of approximately 7 percent. The amendment revises the MPS3 Operating License and Technical Specifications necessary to implement the increased power level.

Date of issuance: August 12, 2008.

Amendment No.: 242.

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

Facility Operating License No. NPF-49: Amendment revised the License and Technical Specifications.

Date of individual notice of issuance in Federal Register : August 20, 2008 (73 FR 49222).

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

Date of amendment request: June 17, 2008.

Brief description of amendment request: The proposed amendment would revise Technical Specification (TS) 5.5.9, Steam Generator (SG) Program, and TS 5.6.9, Steam Generator Tube Inspection Report. For TS 5.5.9, the amendment would incorporate a one-cycle interim alternate repair criteria in the provisions for SG tube Start Printed Page 52422repair criteria during Byron, Unit No. 2, refueling outage 14 and the subsequent operating cycle. For TS 5.6.9, the amendment would revise the current reporting requirements. The proposed changes only affect Byron, Unit No. 2; however, they are docketed for both Byron units because the TSs are common to both units.

Date of publication of individual notice in Federal Register : August 5, 2008 (73 FR 45485).

Expiration date of individual notice: September 5, 2008 (public comment), October 5, 2008 (hearing requests).

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 51.22(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.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

Date of application for amendment: May 10, 2007, as supplemented by letters dated January 10 and July 18, 2008.

Brief description of amendment: The amendment describes the long-term coupon surveillance program for the carborundum samples found in the Unit No. 1 spent fuel pool (SFP). The program verifies that the carborundum degradation rates assumed in the licensee's analyses to prove subcriticality, as required by Title 10 of the Code of Federal Regulations, Section 50.68, remain valid over the 70-year life span of the Unit No. 1 SFP.

Date of issuance: August 27, 2008.

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

Amendment No.: 288.

Renewed Facility Operating License No. DPR-53: Amendment revised the License and fulfills the requirements identified in Appendix C, Additional Conditions, to Renewed Facility Operating License No. DPR-53 as further described in Amendment No. 267 issued on June 3, 2004.

Date of initial notice in Federal Register : June 19, 2007 (72 FR 33780).

The letters dated January 10 and July 18, 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 27, 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 No. 50-461, Clinton Power Station, Unit No. 1, DeWitt 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 Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois; 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, Dauphin County, Pennsylvania

Date of application for amendment: July 19, 2007, as supplemented on July 7, 2008.

Brief description of amendment: The amendments will update the requirements in the Technical Specifications (TS) 5.3.1 “Facility Staff Qualifications,” or TS 6.3.1, “Unit Staff Qualifications,” that have been outdated based on licensed operator training programs accredited by the National Academy for Nuclear Training Academy Document, ACAD 00-003, Revision 1, dated April 2004, and the revised Title 10 of the Code of Federal Regulations, Part 55, “Operators' Licenses.”

Date of issuance: July 25, 2008.

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

Amendment Nos.: 152, 152, 156, 156, 180, 228, 220, 189, 176, 267, 267, 271, 240, 235, 265

Facility Operating License Nos. NPF-72, NPF-77, NPF-37 and NPF-66, NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-16, DPR-55, DPR-56, DPR-29, DPR-30 and DPR-50: The amendments revised the Technical Specifications and License.

Date of initial notice in Federal Register : December 4, 2007 (72 FR 68214). The supplemental letter contained clarifying information, did Start Printed Page 52423not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 25, 2008.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-282, Prairie Island Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota

Date of application for amendment: August 16, 2007, as supplemented by letter dated June 13, 2008.

Brief description of amendment: The amendment revises the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plant, Unit 1. The amendment revises TS 3.8.1 “AC Sources—Operating” to require monthly testing of the Unit 1 emergency diesel generators at or above 2500 kilowatts.

Date of issuance: August 15, 2008.

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

Amendment No.: 187.

Facility Operating License No. DPR-42: Amendment revises the TSs.

Date of initial notice in Federal Register: January 28, 2008 (73 FR 5226).

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

The Commission's related evaluation of the amendment is contained in Safety Evaluation dated August 15, 2008.

No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: August 16, 2007, as supplemented by letter dated June 16, 2008.

Brief description of amendment: The amendment revises the Technical Specification (TS) requirements related to control room envelope habitability in TS 3.7.9, “Control Room Emergency Air Treatment System (CREATS),” and TS Section 5.5, “Programs and Manuals.” The changes are consistent with the Nuclear Regulatory Commission approved Industry/Technical Specification Task Force Traveler No. 448, Revision 3. The availability of this TS improvement was published in the Federal Register on January 17, 2007 (72 FR 2022), as part of the consolidated line item improvement process.

Date of issuance: August 27, 2008.

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

Amendment No.: 105.

Renewed Facility Operating License No. DPR-18: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: October 23, 2007 (72 FR 60035).

The June 16, 2008, supplemental letter 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 27, 2008.

No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket No. 50-362, San Onofre Nuclear Generating Station, Unit 3, San Diego County, California

Date of application for amendments: September 24, 2007, as supplemented by letters dated February 22 and March 27, 2008.

Brief description of amendments: Approves the revision to the SONGS 3 Technical Specification 5.5.2.15, “Containment Leakage Rate Testing Program,” of a one-time extension from the currently approved 15-year interval since the last Integrated Leak Rate Test to a 16-year interval.

Date of issuance: August 15, 2008.

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

Amendment No.: Unit 3-210.

Facility Operating License No. NPF-15: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: October 23, 2007 (72 FR 60036). The supplements dated February 22 and March 27, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission staff original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 15, 2008.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of application for amendment: March 26, 2008.

Brief description of amendment: The proposed amendment would revise the Updated Final Safety Analysis Report (UFSAR) to reflect approval to use the Boiling Water Reactor Vessel and Internals Project reactor pressure vessel integrated surveillance program as the basis for demonstrating the compliance with the requirements of Appendix H to Title 10 of the Code of Federal Regulations Part 50, “Reactor Vessel Material Surveillance Program Requirements.”

Date of issuance: August 14, 2008.

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

Amendment No.: 273.

Renewed Facility Operating License No. DPR-33: Amendment revised the UFSAR.

Date of initial notice in Federal Register: June 3, 2008 (73 FR 31723).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 14, 2008.

No significant hazards consideration comments received: No.

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

Date of application for amendment: August 20, 2007, as supplemented by letter dated March 12, 2008.

Brief description of amendment: The amendment revised Technical Specification 3.8.3, “Diesel Fuel Oil, Lube Oil, and Starting Air,” and its associated Surveillance Requirement 3.8.3.1 to increase the current minimum emergency diesel generator (EDG) fuel oil inventory required to be maintained onsite. The increase in minimum EDG fuel oil would provide conservative margin against potential vortex effects that could occur during fuel oil transfer pump operation.

Date of issuance: August 27, 2008.

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

Amendment No.: 185.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: September 11, 2007 (72 FR 51866). The supplemental letter dated March 12, 2008, provided additional Start Printed Page 52424information 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 as published in the Federal Register.

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.

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 Start Printed Page 52425the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact.[1] Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

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

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

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 ViewerTM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms ViewerTM 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 Start Printed Page 52426depositing 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 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.

Exelon Generation Company, LLC, Docket No. 50-249, Dresden Nuclear Power Station, Unit 3, Grundy County, Illinois

Date of amendment request: August 18, 2008.

Description of amendment request: The amendment revises Technical Specification 3.4.5, “RCS Leakage Detection Instrumentation,” to support implementation of an alternative method of verifying that unidentified leakage in the drywell is within limits.

Date of issuance: August 22, 2008.

Effective date: As of the date of issuance and shall be implemented by 12:00 pm CDT on August 24, 2008.

Amendment No.: 221.

Facility Operating License No. DPR-25: Amendment revises the technical specifications and the operating license.

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

No. On August 17, 2008, the staff issued a Notice of Enforcement Discretion, which was effective immediately and remained in effect until this amendment was issued.

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 22, 2008.

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Generation.

NRC Branch Chief: Russell Gibbs.

Start Signature

Dated at Rockville, Maryland, this 29th day of August 2008.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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.

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[FR Doc. E8-20567 Filed 9-8-08; 8:45 am]

BILLING CODE 7590-01-P