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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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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) 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 Start Printed Page 51327hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from September 10, 2009, to September 23, 2009. The last biweekly notice was published on September 22, 2009 (74 FR 48316).

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 Title 10 of the Code of Federal Regulations (10 CFR), Section 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 and Directives Branch (RDB), TWB-05-B01M, 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 faxed to the RDB at 301-492-3446. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license. 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 O1F21, 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 by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/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, 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.

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other Start Printed Page 51328document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), 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 all adjudicatory 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 an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the petitioner/requestor should contact the Office of the Secretary by e-mail at hearing.docket@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 using the agency's adjudicatory e-filing system 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 Meta-System Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays. The Meta-System Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at MSHD.Resource@nrc.gov.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, 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 request and/or petition 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). 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, unless an NRC regulation or other law requires submission of such information. 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 submissions.

For further details with respect to this license amendment application, 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 O1F21, 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. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov.

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water Reactor, Genoa, Wisconsin (TAC J00359)

Date of amendment request: July 28, 2009.

Description of amendment request: The amendment application proposes changes to Technical Specifications, in support of the dry cask storage project at La Crosse Boiling Water Reactor. The application specifically proposes lower Fuel Element Storage Well water level limits and proposes changes to the definition of “fuel handling.”

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 to the definition of FUEL HANDLING is an administrative Start Printed Page 51329clarification and does not affect the operation of the plant or the postulated accidents in any way. The proposed changes to allow lower Fuel Element Storage Well (FESW) water level limits do not alter the manner in which individual fuel assemblies are moved or alter the design function of the FESW or any other structures, systems, and components used to ensure safe fuel storage. The total number of fuel assembly moves to the Dry Cask Storage System is exactly the same as that contemplated during original plant design when fuel was assumed to be transported from the plant directly to a disposal site. All of the accidents previously evaluated in the La Crosse Boiling Water Reactor (LACBWR) Decommissioning Plan have been reviewed for impact as a result of the proposed water level changes. The proposed changes do not affect the plant in such a manner that the likelihood or consequences of any previously evaluated accident is increased.

Therefore, the proposed amendment 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 to the definition of FUEL HANDLING is an administrative clarification and does not affect the operation of the plant in any way. The proposed changes to allow lower FESW water level limits do not alter the manner in which individual fuel assemblies are moved; or alter the design function of the FESW or any other structures, systems, and components used to ensure safe fuel storage. All of the accidents previously evaluated in the LACBWR Decommissioning Plan have been reviewed for impact as a result of the proposed water level changes. The existing accidents remain applicable and bounding for the LACBWR facility with the proposed changes in place and do not affect the plant in such a manner that a new accident has been created.

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

3. Does the proposed change involve a significant reduction in a margin of safety? No.

The proposed change to the definition of FUEL HANDLING is an administrative clarification and does not affect plant operation or safety margins in any way. The proposed changes to allow lower FESW water level limits do not alter the manner in which individual fuel assemblies are moved; or alter the design function of the FESW or any other structures, systems, and components used to ensure safe fuel storage. All of the accidents previously evaluated in the LACBWR Decommissioning Plan have been reviewed for impact as a result of the proposed water level changes. The likelihood and consequences of previously evaluated accidents remain applicable and bounding with the proposed changes in place; thus, safety margins remain the same.

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

The U.S. 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.

NRC Branch Chief: Andrew Persinko.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: July 23, 2009.

Description of amendment request: The proposed amendment would remove the level indicating instrument from the Technical Specification Surveillance Requirement (SR) for the refueling water storage tank, but leave the low level alarm function in the SR.

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 existing Indian Point 3 Refueling Water Storage Tank (RWST) Technical Specification (TS) Surveillance Requirement (SR) to remove the level indication function for the L-921 instrument loop. Removal of a TS SR for the level indication does not increase the probability of an accident occurring since it is not an accident initiator and does not increase the consequences of an accident since it is not performing any mitigating function and is not a post accident instrument. The proposed revision will not affect RWST lo-lo level alarm function used for operator guidance to begin sequencing to Recirculation Mode of Safety Injection during a postulated loss of coolant accident (LOCA). There will be no change in equipment qualification requirements or changes to the surveillance requirement for the lo-lo level alarm. 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 removes the RWST level indication function from the RWST lo-lo level alarm surveillance requirement for the L-921 instrument loop. The proposed change does not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed change does not result in a change to the way that the equipment or facility is operated so that 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 proposed change removes the RWST level indication function from the RWST io-lo level alarm surveillance requirement for the L-921 instrument loop. There is no change to the design requirements or the surveillance interval. The proposed change does not add the level indicating function elsewhere in the TS because it is a local level indication that is only used during normal operation and was never a post accident monitoring instrument. Therefore the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: July 31, 2009.

Description of amendment request: The proposed change would revise the JAFNPP Technical Specifications (TSs) Surveillance Requirements (SR) for testing of the Residual Heat Removal (RHR) System Shutdown Cooling (SDC) mode Containment Isolation, Reactor Pressure—High Function by replacing the current requirement to perform TS SR 3.3.6.1.3, Perform Channel Calibration, with TS SR 3.3.6.1.1 Perform Channel Check, SR 3.3.6.1.2, Perform Channel Functional Test, SR 3.3.6.1.4, Calibrate the Trip Units, and SR 3.3.6.1.5, Perform Channel Calibration. These changes are to support a proposed plant modification to increase the reliability of SDC isolation logic by changing the source of the reactor high pressure input signal.

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:

Start Printed Page 51330

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

Response: No.

The proposed change modifies the SRs that demonstrate the operability of the SDC Isolation, Reactor Pressure—High function. The current surveillance requirements include a 92-day calibration and a 24-month logic system functional test. These surveillance requirements are typical for pressure switches installed on dedicated process measurement lines. The proposed change in surveillance requirements is consistent with the use of ATTS [Analog Transmitter Trip System] transmitters installed on shared process measurement lines. The proposed surveillance requirements include the standard requirements applied to all ATTS equipment and thus will result in acceptable demonstration of the operability of the SDC Isolation Reactor Pressure—High function.

The ATTS equipment that will be used for the SDC Isolation, Reactor Pressure—High function is classified as safety related and is environmentally qualified. The logic input configuration of the ATTS equipment will be the same as the configuration of the pressure switches. This will assure the same functionality currently performed by the pressure switches currently used for the SDC Isolation Reactor Pressure—High function. The reliability of the ATTS has been proven in other RPS [Reactor Protection System], PCIS [Primary Containment Isolation System], and ECCS [Emergency Core Cooling System] functions and is comparable to the reliability of the pressure switches that currently perform the SDC Isolation, Reactor Pressure—High function. Therefore, the consequences of any accident mitigated by the SDC Isolation, Reactor Pressure—High function will not increase.

Based on these considerations, the proposed surveillance requirement changes do not involve a significant increase in the probability or consequences of an accident 'previously evaluated.

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

Response: No.

The proposed change aligns the TS surveillance requirements with the type of equipment that will be used to supply the reactor pressure input to the SDC Isolation Reactor Pressure—High logic. Since the transmitters that will be used to supply the reactor pressure input are currently installed equipment there are no new accidents introduced by the equipment. The proposed change in SRs aligns the requirements with the—requirements currently imposed on the equipment in other JAF TS applications. The performance of the SDC Isolation, Reactor Pressure—High function, is not altered by changing the input source for reactor pressure parameter. Redundant power sources within the ATTS assure the functionality of the system during all plant operating modes that require the SDC Isolation, Reactor Pressure—High function. The proposed change will not introduce any new failure modes and, therefore, does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The TS surveillance requirements that will be imposed on the SDC Isolation, Reactor Pressure—High function reflect the equipment that will perform that function. The proposed change in surveillance requirements will appropriately demonstrate the operability of the SDC Isolation, Reactor Pressure—High function.

Since the proposed changes to the SRs are consistent with the SRs for ATTS transmitters in other RPS, PCIS, and ECCS applications the proposed requirements have been demonstrated to provide an adequate margin of safety. Therefore, the proposed change does not involve a significant reduction in any margin of safety.

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

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

Date of amendment request: August 5, 2009.

Description of amendment request: Current Technical Specification (TS) 5.5.8, “Inservice Testing Program,” contains references to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI as the source of requirements for the inservice testing (IST) of ASME Code Class 1, 2, and 3 pumps and valves. The proposed amendment would delete the references to Section XI of the Code and incorporate references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). The proposed amendment would also indicate that there may be some nonstandard frequencies utilized in the IST Program in which the provisions of Surveillance Requirement 3.0.2 are applicable. The proposed changes are consistent with Technical Specification Task Force (TSTF) Technical Change Travelers 479-A, “Changes to Reflect Revision to 10 CFR 50.55a,” and 497-A, “Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less.”

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

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

Response: No.

The proposed change revises TS 5.5.8, Inservice Testing Program, for consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are classified as American Society of Mechanical Engineers (ASME) Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code which is consistent with the expectations of 10 CFR 50.55(a).

The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, this proposed change does not represent a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed change does not introduce a new accident initiator, accident precursor, or malfunction mechanism. Therefore, this proposed change does not create the possibility of an accident or a different kind than previously evaluated.

3. 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 revises TS 5.5.8, Inservice Testing Program, for consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code, which is consistent with the expectations of 10 CFR 50.55a. The safety function of the affected pumps and valves are maintained. Therefore, this 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 Start Printed Page 51331amendment request involves no significant hazards consideration.

Attorney for licensee: Terence A. Burke, Associate General Council—Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, Mississippi 39213.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: July 30, 2009.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) Section 3.6.3.1, “Containment Atmosphere Dilution (CAD) System,” to modify containment combustible gas control requirements as permitted by Title 10 of the Code of Federal Regulations, Part 50 Section 50.44 (10 CFR 50.44). 10 CFR 50.44 was revised on September 16, 2003, as noticed in the Federal Register (68 FR 54123).

The Nuclear Regulatory Commission (NRC) staff issued a “Notice Of Opportunity To Comment On Model Safety Evaluation, Model No Significant Hazards Determination, And Model Application For Licensees that Wish To Adopt TSTF-478, Revision 2, `BWR [Boiling-Water Reactor] Technical Specification Changes that Implement the Revised Rule for Combustible Gas Control,” in the Federal Register on October 11, 2007 (72 FR 57970). The notice included a model safety evaluation (SE) and a model no significant hazards consideration (NSHC) determination. On November 21, 2007, the NRC staff issued a notice in the Federal Register (72 FR 65610) announcing that the model SE and model NSHC determination may be referenced in plant-specific applications to adopt the changes. In its application dated July 30, 2009, the licensee affirmed the applicability of the model NSHC determination which is presented below.

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

Criterion 1: The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated

The Containment Atmosphere Dilution (CAD) system is not an initiator to any accident previously evaluated. The TS Required Actions taken when a drywell cooling system fan is inoperable are not initiators to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased.

The revised 10 CFR 50.44 no longer defines a design-basis accident (DBA) hydrogen release and the Commission has subsequently found that the DBA loss-of-coolant accident (LOCA) hydrogen release is not risk significant. In addition, CAD has been determined to be ineffective at mitigating hydrogen releases from the more risk significant beyond DBAs that could threaten containment integrity. Therefore, elimination of the CAD system will not significantly increase the consequences of any accident previously evaluated. The consequences of an accident while relying on the revised TS Required Actions for drywell cooling system fans are no different than the consequences of the same accidents under the current Required Actions. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed 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

No new or different accidents result from utilizing the proposed change. The proposed change permits physical alteration of the plant involving removal of the CAD system. The CAD system is not an accident precursor, nor does its existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building from any design basis event. The changes to the TS do not alter assumptions made in the safety analysis, but reflect changes to the design requirements allowed under the revised 10 CFR 50.44. The proposed change is consistent with the revised safety analysis assumptions.

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

Criterion 3: The proposed change does not involve a significant reduction in a margin of safety

The Commission has determined that the DBA LOCA hydrogen release is not risk significant, therefore is not required to be analyzed in a facility accident analysis. The proposed change reflects this new position and, due to remaining plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, including postulated beyond design basis events, does not result in a significant reduction in a margin of safety.

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

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

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

Attorney for licensee: Mr. J. Bradley Fewell, Associate General Counsel, Exelon Generation Company LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Harold K. Chernoff.

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

Date of amendment request: May 29, 2009.

Description of amendment request: The proposed amendment would: (1) Revise the definition for Operable-Operability in the FCS Technical Specifications (TS); (2) modify the provisions under which equipment may be considered operable when either its normal or emergency power source is inoperable; (3) delete TS limiting condition for operation (LCO) 2.0.1(2); (4) delete diesel generator surveillance requirement (SR) 3.7(1)e; and (5) relocate the guidance for inoperable power supplies and verifying operability of redundant components into the LCO for electrical equipment 2.7, Electrical Systems.

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 change to revise the definition of operable-operability, modify the provisions under which equipment may be considered operable when either its normal or emergency power source is inoperable, delete Technical Specification (TS) limiting conditions for operation (LCO) 2.0.1(2), and relocate the guidance for inoperable power supplies and verifying operability of redundant components into the LCO for electrical equipment is more aligned with NUREG-1432, Standard Technical Specifications [STS] for Combustion Engineering Plants, and does not adversely impact the probability of an accident Start Printed Page 51332previously evaluated. The proposed changes are being made to address inconsistencies in guidance provided in TS 2.0.1(2) and TS 2.7(2). The proposed change does not affect the operability requirements for the emergency diesel generators (EDGs) or the house service transformers, and therefore does not impact the consequences of an analyzed accident.

The new requirement added to TS 2.7 provides assurance that a loss of offsite power during the period that an EDG (or house service transformer) is inoperable, or loss of an EDG during the period that a house service transformer is inoperable, or loss of a house service transformer during the period that an EDG is inoperable, does not result in a complete loss of safety function of critical systems; thereby such a loss does not significantly increase the probability of an accident.

Consistent with NUREG 1432, the 4-hour allowed time added to TS 2.7(2)j for the EDGs, takes into account the capacity and capability of the remaining alternating current (AC) sources, a reasonable time for repairs, and the low probability of a design basis accident (DBA) occurring during this period. On a component basis, single failure protection for the required feature's function may have been lost; however, function has not been lost.

Additionally, consistent with NUREG-1432, the 24-hour allowed time added to TS 2.7(2)b for the house service transformers takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period.

The proposed change removes the surveillance requirement (SR) to perform an inspection of the EDG on a refueling inspection frequency in accordance with the manufacturer's recommendations. This inspection is considered a maintenance activity, not an SR, and has no impact on the probability of an accident since EDGs are not initiators for any analyzed event. Deletion of TS SR 3.7(1)e from the TS does not impact the capability of the EDGs to perform their accident mitigation functions. The required EDG maintenance inspections will continue to be performed in accordance with the licensee-controlled EDG maintenance process. The consequences of an accident are not impacted because EDG operability is controlled by other portions of TS 3.7, which ensures that required surveillances are performed. The appropriate LCOs are entered in the event that EDG surveillance criteria are not met.

As a result of redefining “OPERABLE” and adding the provision to TS 2.7(2)j, the statements “provided there are no inoperable required engineered safeguards components which are redundant” related to the electrical distribution components are being deleted from the other 2.7(2) TS for the buses, transformer, and motor control center (MCC) for clarification and consistency because these statements restrict only to engineered safeguards components. In addition, the administrative changes to renumber the existing TS sections “TS 2.0.1(3) to 2.0.1(2)” and TS 3.7(1)f to TS 3.7(1)e. are being made as a result of deletions to previous TS paragraphs and are being made for consistency and clarification. Rearranging the listing order of the MCCs in TS 2.7(1)f and TS 2.7(2)g in bus order clarifies the TS. As such, these editorial changes are not initiators of any accidents previously evaluated. As a result, the probability of an accident previously evaluated is not affected.

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.

The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change in methods governing normal plant operation. The proposed changes to TS 2.0.1(2) and TS 2.7 do not create the possibility of a new or different kind of accident since the design function of the affected equipment is not changed. No new interactions between systems or components are created. No new failure mechanisms of associated systems will exist.

By deleting TS LCO 2.0.1(2) and including the guidance in TS 2.7, inconsistencies in the existing TS will be eliminated. The new requirements added to TS 2.7 will include guidance to declare required systems or components without a normal or emergency power source available inoperable, when a redundant system or component is also inoperable. This provides assurance that a loss of offsite power, during the period that an EDG (or house service transformer) is inoperable, or loss of an EDG during the period that a house service transformer is inoperable (or vice versa), does not result in a complete loss of safety function of critical systems.

No new failure mechanisms would be created. The proposed changes do not alter any assumptions made in the safety analyses. For the most part, the proposed changes are more aligned with the STS.

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.

The proposed changes to delete TS 2.0.1(2) and relocate the guidance for inoperable power supplies and verifying operability of redundant components to TS LCO 2.7(2)j, to delete the statement that MCC-3C1 may be inoperable in excess of 8 hours if battery chargers No. 1 and No. 2 are operable, and to delete the SR for inspecting the DG on a refueling frequency in accordance with the manufacturer's recommendations do not alter the manner in which safety limits or limiting safety system settings are determined. The safety analysis acceptance criteria are not affected by these proposed changes. The sources of power credited for design basis events are not affected by the proposed changes.

The proposed changes to modify the provisions under which equipment may be considered operable when either its normal or emergency power source is inoperable, delete TS LCO 2.0.1(2), and relocate the guidance for inoperable power supplies and verifying operability of redundant components into the LCO for electrical equipment is more aligned with the STS. These changes are being made to address inconsistencies in guidance provided in TS 2.0.1(2) and TS 2.7(2). The proposed change does not reduce the operability requirements for the transformers, buses, MCCs, or EDGs and therefore will not result in plant operation in a configuration outside of the design basis.

Further, the proposed change does not change the design function of any equipment assumed to operate in the event of an accident.

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: David A. Repka, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: March 24, 2009, as supplemented by letters dated April 24, and September 11, 2009.

Description of amendment request: The proposed change revises the allowable value in the Technical Specification (TS) Table 3.3.5.1-1 (Function 3.d) for the high-pressure coolant injection (HPCI) automatic pump suction transfer from the condensate storage tank (CST) to the suppression pool (SP). The present allowable value for this transfer is greater than or equal to 36 inches above the CST bottom. The proposed change is to increase the allowable value for this transfer to occur at greater than or equal to 40.5 inches above the CST bottom.

Additionally, the proposed amendment also includes an editorial/administrative change which corrects a typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f.

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:

Start Printed Page 51333

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

No. The proposed change to TS Table 3.3.5.1-1 increases the Technical Specification allowable value for the HPCI suction low level automatic transfer function from ≥ 3 6 inches to ≥ 40.5 inches above the CST bottom. There are no process setpoint changes associated with this TS allowable value change. This TS change does not introduce the possibility of an increase in the probability or consequences of an accident because the HPCI automatic transfer function is not an initiator of any new accidents nor does it introduce any new failure modes. The CST is not safety related and therefore not credited in any design basis accident analyses. However, the CST reserve volume is credited in anticipated transients without scram (ATWS), Appendix R and station blackout (SBO) evaluations. The reserve volume available in the CST at the proposed allowable value of 40.5 inches above the CST bottom remains adequate to fully support these HPCI system support functions and the change fully supports HPCI system operation. The reserve volume is not reduced as a result of the proposed change in the TS allowable value since the transfer will still occur at the CST low level instrument setpoint of 43.5 inches above tank bottom, which remains unchanged.

The HPCI system automatic transfer function occurs at the point in a design basis accident (DBA) when the CST level reaches the low level transfer setpoint. This proposed change will require the HPCI pump suction to be transferred from the CST to the SP at 40.5 inches versus 36 inches above the CST bottom. Currently, the TS allow this transfer to occur at 36 inches. This proposed change is conservative because it assures the suction transfer will occur while there is more water in the tank, thus eliminating the possibility of vortex formation and air intrusion to the HPCI pump suction. Since this proposed change ensures the HPCI system automatic suction transfer function occurs without adversely impacting HPCI system operation, it does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed editorial/administrative change is necessary to correct a typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f. This editorial 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. As discussed above, the proposed change to TS Table 3.3.5.1-1 involves increasing the TS allowable value for the HPCI low level automatic transfer function from the CST to the SP at ≥ 36 inches to ≥ 40.5 inches above the CST tank bottom. This change ensures the HPCI automatic transfer function occurs without introducing the possibility of vortex formation or air intrusion in the HPCI pump suction path. All HPCI system support functions remain unaffected by this change. This TS change does not introduce the possibility of a new accident because the HPCI automatic transfer function is not an initiator of any accident and no new failure modes are introduced. There are no new types of failures or new or different kinds of accidents or transients that could be created by these changes. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed editorial/administrative change only corrects a typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f. This editorial 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 margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change to TS Table 3.3.5.1-1 involves increasing the allowable level at which the HPCI automatic suction transfer from the CST to the SP must occur to avoid the possibility of vortex formation or air intrusion into the HPCI pump. This change does not result in a change to the level switch setpoint, which initiates the HPCI suction transfer from the CST to the SP. Although the allowable value for the transfer is now closer to the process setpoint for activation of the level switch, this reduction in operating margin was reviewed and determined to be acceptable. The level switch setpoint tolerances were established based on historical instrument data and instrument characteristics. These tolerances provide adequate margin to the proposed TS allowable value of 40.5 inches above the CST bottom. The tolerances further ensure the transfer will occur prior to level reaching the technical specification allowable value. Therefore, the proposed change does not result in a significant reduction in a margin of safety.

The proposed editorial/administrative change only corrects a typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f. This editorial change does not result in a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: July 30, 2009.

Description of amendment request: The proposed amendment would relocate Technical Specification (TS) surveillance requirements (SRs) for the reactor recirculation system motor-generator (MG) set scoop tube stop settings to the Technical Requirements Manual (TRM). Specifically, the proposed amendment would relocate TS SR 4.4.1.1.3 to the TRM which is a licensee-controlled document. SR 4.4.1.1.3 requires that each MG set scoop tube mechanical and electrical stop be demonstrated operable with overspeed setpoints less than or equal to 109% and 107%, respectively, of rated core flow, at least once per 18 months.

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. The Nuclear Regulatory Commission (NRC) staff's review 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 major components in the MG set consist of a motor, fluid coupler and a generator. The motor drives the generator through the fluid coupler. The speed and output of the generator rise and fall as the volume of fluid in the coupler is varied by changing the position of the scoop tube. As the generator's output increases or decreases, the speed of the recirculation pump follows suit. The scoop tube mechanism has both mechanical and electrical overspeed stops that limit recirculation flow by limiting the MG set speed. The electrical stop actuates first. The mechanical stop is designed to prevent the scoop tube motion if the electrical stop fails or to mitigate overshoot of the electrical stop. The electrical stops are not credited in any of the accident or transient analyses. The mechanical stop settings are an input used in the determination of the flow dependent minimum critical power ratio (MCPR) and the linear heat generation rate (LHGR) or average planar linear heat generation rate (APLHGR) operating limits. These operating limits are established and documented on a cycle-specific basis in the core operating limits report (COLR) in accordance with TS 6.9.1.9. Operation within the MCPR, LGHR and APLHGR operating limits is required in accordance with TSs 3.2.3, 3.2.4, and 3.2.1, respectively.

Once relocated, any future changes to the surveillance requirements for the MG set scoop tube mechanical and electrical stop settings would be controlled by 10 CFR 50.59.

There are no physical plant modifications associated with this change. The proposed amendment would not alter the way any structure, system, or component (SSC) functions and would not alter the way the plant is operated. As such, the proposed Start Printed Page 51334amendment would have no impact on the ability of the affected SSCs to either preclude or mitigate an accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment would not change the design function or operation of the SSCs involved and would not impact the way the plant is operated. As such, the proposed change would not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases. 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 margin of safety is associated with the confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant pressure boundary, and containment structure) to limit the level of radiation to the public. There are no physical plant modifications associated with the proposed amendment. The proposed amendment would not alter the way any SSC functions and would not alter the way the plant is operated. The proposed amendment would not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. The proposed amendment would have no impact on the structural integrity of the fuel cladding, reactor coolant pressure boundary, or containment structure. Based on the above considerations, the NRC staff concludes that the proposed amendment would not degrade the confidence in the ability of the fission product barriers to limit the level of radiation to the public. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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: Vincent Zabielski, PSEG Nuclear LLC-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: Harold K. Chernoff.

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 System (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.resource@nrc.gov.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of application for amendments: May 21, 2009.

Brief description of amendments: The amendments removed the Table of Contents from the Technical Specifications and place them under licensee control.

Date of issuance: September 21, 2009.

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

Amendment Nos.: 293 and 269.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: June 30, 2009 (74 FR 31320).

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated September 21, 2009.

No significant hazards consideration comments received: No.

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

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

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

Date of application for amendments: February 27, 2009.

Brief description of amendments: The amendments deleted those portions of the Technical Specifications (TSs) superseded by the Code of Federal Regulations, Part 26, Subpart I. The changes are consistent with Nuclear Regulatory Commission (NRC)-approved Revision 0 to Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF-511, “Eliminate Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.”

Date of issuance: September 21, 2009.

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

Amendment Nos.: 251 and 246.

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

Amendment Nos.: 253 and 233.

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

Amendment Nos.: 365, 367, and 366.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and technical specifications.

Date of initial notices in Federal Register: August 11, 2009 (74 FR 40236) Catawba and McGuire; and August 11, 2009 (74 FR 40237) Oconee.

The Commission's related evaluation and final finding of no significant hazards consideration of the Start Printed Page 51335amendments is contained in a Safety Evaluation dated September 21, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 13, 2008, as supplemented by letters dated June 1, July 14, and August 17, 2009.

Brief description of amendment: The amendment modified Technical Specification 3.3.1.1, Reactor Protective Instrumentation, specifically Table 4.3-1 and associated Notes 7 and 8, to clarify and streamline Reactor Coolant System flow verification requirements associated with the Departure from Nucleate Boiling Ratio reactor trip signal.

Date of issuance: September 16, 2009.

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

Amendment No.: 286.

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

Date of initial notice in Federal Register: January 27, 2009 (74 FR 4769). The supplemental letters dated June 1, July 14, and August 17, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 16, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 13, 2009, as supplemented by letter dated July 8, 2009.

Brief description of amendment: The amendment modified Technical Specification 2.1.1.1, departure from nucleate boiling ratio safety limit based upon the Combustion Engineering 16 $× 16 Next Generation Fuel design and the associated departure from nucleate boiling correlations.

Date of issuance: September 18, 2009.

Effective date: As of the date of issuance and shall be implemented after the current cycle (Cycle 20) is completed and prior to startup for operating Cycle 21.

Amendment No.: 287.

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

Date of initial notice in Federal Register: June 30, 2009 (74 FR 31321). The supplemental letter dated July 8, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 18, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 25, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082110187), as supplemented by letters dated October 31, 2008 (ADAMS Accession No. ML083080059), February 17, 2009 (ADAMS Accession No. ML090480372), May 8, 2009 (ADAMS Accession No. ML092380433) and July 27, 2009 (ADAMS Accession No. ML092100162).

Brief description of amendments: The amendments revised Technical Specification (TS) 3.3.1.1, “Reactor Protection System (RPS) Instrumentation,” Surveillance Requirement (SR) 3.3.1.1.8 and TS 3.3.1.3, “Oscillation Power Range Monitor (OPRM) Instrumentation,” SR 3.3.1.3.2 to increase the frequency interval between Local Power Range Monitor calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH.

Date of issuance: September 16, 2009.

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

Amendment Nos.: 195/182.

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

Date of initial notice in Federal Register: January 23, 2009 (74 FR 4250-4251). The October 31, 2008, February 17, 2009, May 8, 2009, and July 27, 2009 supplements, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration nor expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 16, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 9, 2008, supplemented by letter dated April 2, 2009.

Brief description of amendment: The amendment reflects the planned installation of replacement steam generators (SGs). Specifically, the amendment modified the technical specifications to eliminate the existing requirements associated with tube sleeve repairs and alternate repair criteria which are not applicable to the replacement SGs. It also incorporated a revised primary-to-secondary leakage criterion, changes the required reporting period for SG inspection results, and incorporated revised tube integrity surveillance frequency requirements to reflect the new Alloy 690 tubing material.

Date of issuance: September 15, 2009.

Effective date: Upon installation of the replacement SGs and shall be implemented prior to exiting cold shutdown from the TMI-1 SG replacement refueling outage (T1R18), which is scheduled to begin in the fall of 2009.

Amendment No.: 271.

Facility Operating License No. DPR-50: Amendment revised the license and the technical specifications.

Date of initial notice in Federal Register: March 10, 2009 (74 FR 10310).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 2009.

No significant hazards consideration comments received: No.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of application for amendment: January 30, 2009, as supplemented by a letter dated July 30, 2009.

Brief description of amendment: The amendment deleted the Duane Arnold Energy Center (DAEC) Technical Specification (TS) Section 5.2.2.e regarding work hour controls.

Date of issuance: September 18, 2009.

Effective date: As of the date of issuance and shall be implemented by October 1, 2009. Amendment No.: 274.Start Printed Page 51336

Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 24, 2009 (74 FR 12393). The supplemental letter contained clarifying information, did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 18, 2009.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 13, 2008, as supplemented by letters dated April 8, May 29, June 12, and September 1, 2009.

Brief description of amendment: The amendment revised the licensing basis by approving adoption of the Alternative Source Term (AST), in accordance with Section 50.67 of Title 10 of the Code of Federal Regulations (10 CFR), for use in calculating the loss-of-coolant accident (LOCA) dose consequences. The amendment revised the Technical Specifications (TSs) to (1) change the TS definition for DOSE EQUIVALENT I-131 to adopt Federal Guidance Report 11 dose conversion factors; (2) require operability of the Standby Liquid Control system in Mode 3, to reflect its credit in the LOCA analysis; (3) establish a Main Steam (MS) Pathway leakage limit that effectively increases the previous MS isolation valve leakage limit; and (4) change TS Section 5.5.12 to reflect a requested permanent exemption from the requirements of 10 CFR Part 50, Appendix J, Option B, Section III.A, to allow exclusion of MS Pathway leakage from the overall integrated leakage rate measured during the performance of a Type A test, and from the requirements of Appendix J, Option B, Section III.B, to allow exclusion of the MS Pathway leakage from the combined leakage rate of the penetrations and valves subject to Type B and C tests.

Date of issuance: September 15, 2009.

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

Amendment No.: 234.

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

Date of initial notice in Federal Register: January 23, 2009 (74 FR 4251). The supplemental letters dated April 8, May 29, June 12, and September 1, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 2009.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 30, 2009.

Brief description of amendment request: The amendments revised Technical Specification (TS) by deleting the Reactor Coolant Pump breaker position reactor trip in TS 3.3.1, “Reactor Trip System (RTS) Instrumentation.”

Date of Issuance: September 18, 2009.

Amendment Nos.: Unit 1-183; Unit 2-176.

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

Date of initial notice in Federal Register : May 19, 2009 (74 FR 23448).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 18, 2009.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 5, 2009.

Brief description of amendment: The amendment revised WBN Unit 1 technical specifications (TSs) to revise the completion time from 1 hour to 24 hours for Condition B of TS 3.5.1, “Accumulators” and its associated Bases.

Date of issuance: September 9, 2009.

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

Amendment No.: 81.

Facility Operating License No. NPF-90: Amendment revises TS 3.5.1.

Date of initial notice in Federal Register : June 30, 2009 (74 FR 31326).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 9, 2009.

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 Start Printed Page 51337issued 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 O1F21, 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 email to pdr.resource@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, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license. 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 O1F21, 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.resource@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

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

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

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

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

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 Start Printed Page 51338hearing. 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.

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), 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 all adjudicatory 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 hearing.docket@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 website 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 using the agency's adjudicatory e-filing system 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 Meta-System Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays. The Meta-System Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at MSHD.Resource@nrc.gov.

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).

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, unless an NRC regulation or other law requires submission of such information. 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.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

Date of application for amendment: September 3, 2009, as supplemented on September 8, 2009.

Brief description of amendment: The amendment revised the Diablo Canyon Power Plant, Unit No. 2 Technical Specification (TS) 3.7.1, “Main Steam Safety Valves (MSSVs),” by increasing the Power Range Neutron Flux High setpoint in TS Table 3.7.1-1 from 87 percent rated thermal power (RTP) to 106 percent RTP. This will allow the unit to operate at full power with one main steam safety valve, MS-2-RV-224, inoperable for the remainder of Cycle 15.

Date of issuance: September 17, 2009.

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

Amendment No.: 208.

Facility Operating License No. DPR-82: The amendment revised the Facility Operating License and Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. A public notice of the proposed amendment was published in The Tribune newspaper, located in San Luis Obispo, California, Start Printed Page 51339on September 11 and 12, 2009. The notice provided an opportunity to submit comments on the NRC staff's proposed NSHC determination.

The supplemental letter dated September 8, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in The Tribune.

The Commission's related evaluation of the amendment, finding of exigent circumstances, consideration of public comments, state consultation, and final NSHC determination are contained in a safety evaluation dated September 17, 2009.

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: Michael T. Markley.

Start Signature

Dated at Rockville, Maryland, this 25th day of September 2009.

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. E9-23780 Filed 10-5-09; 8:45 am]

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