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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 (NRC or the Commission) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from November 3, 2010, to November 17, 2010. The last biweekly notice was published on November 16, 2010 (75 FR 70032).

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), 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, Rules, Announcements and Directives Branch (RADB), 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 RADB 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, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.

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, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20854. Publicly available records will be accessible from the Agencywide Start Printed Page 74092Documents 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 requestor/petitioner 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 requestor/petitioner 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 requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/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. 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 requestor/petitioner to relief. A requestor/petitioner 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 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 (72 FR 49139, August 28, 2007). 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 participant should contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by telephone at 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 (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

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. System requirements for accessing the E-Submittal server are detailed in NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant 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 documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing 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 E-Filing 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 Start Printed Page 74093apply 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 by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html, by e-mail at MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

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, or the 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.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

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 e-mail to pdr.resource@nrc.gov.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: September 24, 2010.

Description of amendment request: The proposed amendment would revise the Fermi 2 Radiological Emergency Response Preparedness (RERP) Plan to increase the staff augmentation times for Technical Support Center-related functions from 30 to 60 minutes and for Emergency Operations Facility-related functions from 60 to 90 minutes.

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

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

The proposed extension of staff augmentation times has no effect on normal plant operation or on any accident initiator. The change affects the response to radiological emergencies under the Fermi 2 Radiological Emergency Response Preparedness (RERP) Plan. The ability of the emergency response organization to respond adequately to radiological emergencies has been evaluated. Improvements have been made to equipment, procedures, and training since initial approval of the Fermi 2 Emergency Plan that have resulted in a significant increase in the on-shift capabilities and knowledge such that there would be no degradation or loss of Emergency Plan function as a result of the proposed change. A functional analysis was also performed on the effect of the proposed change on the timeliness of performing major tasks for the major functional areas of the RERP Plan. The analysis concluded that extension of staff augmentation times would not significantly affect the ability to perform the required tasks.

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

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

The proposed change affects the required response times for supplementing onsite personnel in response to a Radiological emergency. It has been evaluated and determined not to significantly affect the ability to perform that function. It has no effect on the plant design or on the normal operation of the plant and does not affect how the plant is physically operated under emergency conditions. The extension of staff augmentation times in the RERP Plan does not affect the plant Operating, Abnormal Operating, or Emergency Operating procedures which are performed by plant staff during all plant conditions.

Therefore, since the proposed change does not affect the design or method of operation of the plant, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change does not affect plant design or method of operation. 10 CFR 50.47(b) and 10 CFR Part 50, Appendix E establish emergency planning standards that require adequate staffing, satisfactory performance of key functional areas and critical tasks; and timely augmentation of the response capability. Since the initial NRC approval of the Emergency Plan, there have been improvements in the technology used to support the RERP functions and in the capabilities of onsite personnel. A functional analysis was performed on the effect of the proposed change on the timeliness of performing major tasks for the functional areas of the RERP Plan. The analysis concluded that an increase in staff augmentation times would not significantly affect the ability to perform the required RERP tasks. Thus, the proposed change has been determined not to adversely affect the ability to meet the emergency planning standards as described in 10 CFR 50.47(b) and 10 CFR Part 50, Appendix E.

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

The NRC staff has reviewed the licensee's analysis and, based on this Start Printed Page 74094review, 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 G. Pettinari, Attorney—Corporate Matters, 688 WCB, Detroit Edison Company, One Energy Plaza, Detroit, Michigan 48226-1279.

NRC Branch Chief: Robert J. Pascarelli.

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: January 6, 2010, as supplemented by letters dated August 20, 2010, and October 14, 2010.

Description of amendment request: The proposed amendment would enable PBAPS, Units 2 and 3, to possess byproduct and special nuclear material from Limerick Generating Station (LGS), Units 1 and 2. Specifically, the revised license paragraph would permit storage of low-level radioactive waste (LLRW) from LGS in the PBAPS LLRW Storage Facility. The PBAPS LLRW Storage Facility currently provides storage for LLRW generated at PBAPS.

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 (NSHC), which is presented below:

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

Response: No.

The proposed change is an amendment to the PBAPS FOLs [Facility Operating Licenses] that will enable PBAPS to receive and store Class B/C LLRW from LGS in the PBAPS LLRWSF [Low Level Radioactive Waste Storage Facility]. This proposed change does not impact any initiators or precursors of previously analyzed accidents. The storage of Class B/C LLRW from LGS does not impact the failure of any plant structures, systems, or components. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed change does not affect any active or passive failure mechanisms that could lead to an accident. The PBAPS LLRWSF is not safety related, and is not used for plant shutdown resulting from accident or nonstandard operational conditions.

The proposed change does not significantly increase the consequences of postulated design basis events (i.e., seismic, flood, tornado, fire, and container drop events), in that the postulated impact of these events remains well below regulatory requirements (i.e., less than 10 percent of 10 CFR Part 100, “Reactor Site Criteria” acceptance criteria).

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 previously evaluated?

Response: No.

The proposed change is an amendment to the PBAPS FOLs that will enable PBAPS to receive and store Class B/C LLRW from LGS in the PBAPS LLRWSF. The proposed amendment does not involve any change to the plant equipment or system design functions. EGC has verified that the storage of Class B/C LLRW from LGS in the PBAPS LLRWSF does not affect the ability of the PBAPS LLRWSF to perform its design function, including compliance with NRC regulatory requirements and guidance. No new accident initiators are introduced by this amendment.

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

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

Response: No.

The proposed change is an amendment to the PBAPS FOLs that will enable PBAPS to receive and store Class B/C LLRW from LGS in the PBAPS LLRWSF. The proposed amendment does not involve any change to plant equipment or system design functions. The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the setpoints for the actuation of equipment relied upon to respond to an event. The proposed amendment does not affect the PBAPS safety limits or setpoints at which protective actions are initiated.

The proposed amendment does not significantly increase the dose rate at the exterior wall of the LLRWSF, the nearest restricted area boundary, and the nearest residence when the LLRWSF is filled to capacity with Class B/C LLRW. Therefore, these dose rates will remain within limits specified in 10 CFR Part 20 and 40 CFR Part 190.

Additionally, the potential radiological impact of a postulated design basis container drop accident is less than 10 percent of the 10 CFR Part 100 acceptance criteria.

Therefore the margin of safety is not reduced by the proposed change.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC.

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.

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: June 25, 2010, as supplemented by letter dated August 16, 2010.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3, “Primary Containment Isolation Valves (PCIVs),” and SR 3.6.1.5, “Reactor Building-to-Suppression Chamber Vacuum Breakers,” to modify the required level for the liquid nitrogen storage tank.

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 (NSHC), 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 TS changes to increase the level in the liquid nitrogen storage tank from ≥ 16 inches water column to a level of ≥ 22 inches water column, or equivalent volume of ≥ 124,000 scf [standard cubic feet] at 250 psig, is necessary in order to correct a non-conservative TS value. Increasing the level is intended to ensure continued operability of the PCIVs (SR 3.6.1.3.1) and Reactor Building-to-Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) via the SGIG [safety grade instrument gas] system. The non-conservative TS condition was identified based on a re-analysis of the liquid nitrogen storage tank operation. The leakage allowance that was previously assumed was not based on a rigorous empirical value. The re-analysis of the leakage allowance assumes more reasonable system leakage based on operational data. Exelon determined that the current PBAPS, Units 2 and 3, TS SR value for the minimum level in the liquid nitrogen storage tank of ≥ 16 inches water column is non-conservative and that the guidance of Nuclear Regulatory Commission (NRC) Administrative Letter 98-10, “Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety,” applies. Exelon has implemented administrative controls to maintain the amount of nitrogen in the liquid nitrogen storage tank at a level of > 22 inches water column in support of SGIG system operation.

Exelon is submitting this License Amendment Request to address this non-conservative condition. The proposed TS Start Printed Page 74095changes do not introduce new equipment or new equipment operating modes, nor do the proposed changes alter existing system relationships. The proposed changes do not affect plant operation, design function or any analysis that verifies the capability of a system, structure or component (SSC) to perform a design function. Further, the proposed changes do not increase the likelihood of the malfunction of any SSC or impact any analyzed accident. Consequently, the probability or consequences of an accident previously evaluated are not affected.

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

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

Response: No.

The proposed TS change to increase the level in the liquid nitrogen storage tank from ≥ 16 inches water column to a level of ≥ 22 inches water column, or equivalent volume of ≥ 124,000 scf at 250 psig, for the PCIVs (SR 3.6.1.3.1) and Reactor Building-to-Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) is needed to correct a non-conservative value based on a revised analysis. The proposed TS changes do not alter the design function or operation of any SSC. There is no new system component being installed, no construction of a new facility, and no performance of a new test or maintenance function. The proposed TS changes do not create the possibility of a new credible failure mechanism or malfunction. The proposed changes do not modify the design function or operation of any SSC. Further, the proposed changes do not introduce new accident initiators. Consequently, the proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

The proposed TS changes to increase the level in the liquid nitrogen storage tank from ≥ 16 inches water column to a level of ≥ 22 inches water column, or equivalent volume of ≥ 124,000 scf at 250 psig, for the PCIVs (SR 3.6.1.3.1) and Reactor Building-to-Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) are necessary to correct an existing non-conservative TS value. The proposed TS changes are needed based on a revised analysis that utilizes empirical data for nitrogen system uses and losses. The proposed changes do not exceed or alter a design basis or a safety limit for a parameter established in the PBAPS, Units 2 and 3, Updated Final Safety Analysis Report (UFSAR) or the PBAPS, Units 2 and 3, Renewed Facility Operating License (FOL). Consequently, the proposed changes do not result in a reduction in the margin of safety.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC.

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.

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

Date of amendment request: September 22, 2010.

Description of amendment request: The proposed amendment would relocate the list of pumps, fans, and valves in Technical Specification (TS) 4.5.1.1b, Sequence and Power Transfer Test, to the Three Mile Island, Unit 1 (TMI-1) Updated Final Safety Analysis Report. In addition, TS 4.5.1.2b, TS 4.5.2.2a, and TS 4.5.2.2b refer to this test and are proposed for revision to reflect the proposed change to TS 4.5.1.1b.

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 with an NRC edit in brackets:

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

Response: No.

The proposed license amendment does not add, delete or modify plant equipment. The proposed changes are administrative in nature. The proposed amendment would relocate the list of pumps, fans and valves in Technical Specification (TS) 4.5.1.1b, Sequence and Power Transfer Test, to the TMI-1 Updated Final Safety Analysis Report (UFSAR) Section 8.3, Tests and Inspections.

The proposed changes relocate surveillance requirement details that are not required by 10 CFR 50.36, and are [partially] consistent with standard technical specifications, NUREG-1430, “Standard Technical Specifications Babcock and Wilcox Plants.” The proposed changes do not change current surveillance requirements. The subject list of pumps, fans and valves that will be relocated to the UFSAR Section 8.3 will continue to be administratively controlled and future changes will be controlled under 10 CFR 50.59.

The probability of an accident is not increased by these proposed changes because the Sequence and Power Transfer Test is not an initiator of any design basis event. Additionally, the proposed changes do not involve any physical changes to plant structures, systems, or components (SSCs), or the manner in which these SSCs are operated, maintained, or controlled. The consequences of an accident will not be increased because the proposed administrative changes to the Sequence and Power Transfer Test and Sequence Test will continue to provide a high degree of assurance that the Electric Power System will meet its safety related function.

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

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

Response: No.

The proposed changes do not alter the physical design, safety limits, safety analyses assumptions, or the manner in which the plant is operated or tested. The proposed changes are administrative in nature and the surveillance requirements remain the same. Accordingly, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant SSC in the performance of their safety function.

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

3. Does the proposed amendment involve a significant reduction in the 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 changes to SSCs or operating and testing procedures associated with the proposed amendment.

The proposed changes do not impact the assumptions of any design basis accident, and do not alter assumptions relative to the mitigation of an accident or transient event. The proposed changes are administrative in nature and the surveillance requirements remain the same.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, with the NRC edit noted above incorporated, 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: J. Bradley Fewell, Esquire, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Harold K. Chernoff.Start Printed Page 74096

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

Date of amendment request: September 24, 2010.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.4.1.2.3, to allow up to two Main Steam Safety Valves (MSSVs) per steam generator to be inoperable with no required reduction in power level. It would also revise the required maximum overpower trip setpoints for any additional inoperable MSSVs consistent with the plant transient analysis. The proposed change requires that with less than four MSSVs associated with either steam generator operable, the plant would be required to be brought to the hot shutdown condition.

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, with NRC edits in brackets:

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

Response: No.

The proposed amendment is not a change to the plant structures, systems, or components. There is no increase to the likelihood of Main Steam Safety Valve (MSSV) related failures. The MSSVs are relied upon to mitigate the effects of Updated Final Safety Analysis Report (UFSAR) Chapter 14 design basis events including the loss of load (turbine trip), which is the limiting event for secondary system overpressure. Analyses, performed in accordance with NRC approved methods, have demonstrated that with reduced MSSV availability and following the specified power level restrictions, the MSSVs will continue to limit the secondary system pressure to less than 110 percent of the design pressure of the Once Through Steam Generators (OTSGs) and the Main Steam (MS) System as required by [the American Society of Mechanical Engineers] ASME code. 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 previously evaluated?

Response: No.

The proposed amendment is not a change to the plant structures, systems, or components (SSCs). Furthermore, within the current licensing basis, the MSSVs are accident mitigation SSCs. The current licensing basis does not [explicitly] include consideration of a MSSV failure as an event initiator [and a failed open MSSV has been shown to be bounded by the larger maximum break size analysis presented in the TMI-1 UFSAR]. The proposed amendment will not fundamentally alter or create any new operator actions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The limiting event for secondary system overpressure is a loss of load event (turbine trip). The event has been analyzed for varying MSSVs out of service, using NRC approved methods. The results of the analysis demonstrate that the existing design acceptance criteria (i.e., MS and OTSG pressure remain less than 110 percent of the design pressure) are met for all combinations of inoperable MSSVs and initial power levels described in the proposed change. 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, including the edits listed above, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Harold K. Chernoff.

FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of amendment request: January 27, 2010.

Description of amendment request: The proposed changes would amend Renewed Facility Operating Licenses DPR-24 and DPR-27 for the Point Beach Nuclear Plant, Units 1 and 2, respectively. The proposed amendment consists of changes to Technical Specification 3.8.3, “Diesel Fuel Oil and Starting Air.”

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

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

Response: No.

This proposed change increases the minimum required amount of stored diesel fuel in the associated common fuel oil storage tank [FOST] for two standby emergency power sources to start, load to their respective loading limits and to operate continuously up to a maximum of 48 hours. An increase in the minimum required fuel oil volume required in the fuel oil storage tanks does not increase the probability or consequences of an accident previously evaluated.

[Limiting Condition for Operation] LCO 3.8.3 Condition A, currently requires that one or more standby emergency power sources have ≥ 11,000 gallons of fuel when the associated [emergency diesel generator] EDG is declared operable. The proposed change increases the amount of stored fuel to ≥ 24,000 gallons for two standby EDGs. It further adds new Required Action A.2 if the FOST stored capacity falls below the minimum required values. The proposed change also accounts for instrument indicator loop uncertainty values for unusable volume.

New LCO [3.8.3] Condition B, addresses the case of one EDG operating in either Train “A” or Train “B.” The new condition specifies that the minimum volume of diesel fuel required to support continued operation of a single EDG for 48 hours at rated load is ≥ 13,000 gallons. This proposed change also accounts for instrument indicator loop uncertainty values for unusable volume.

[Surveillance Requirement] SR 3.8.3.1 is revised to reflect the increased amount of diesel fuel required to be maintained to support operation of the EDGs following recalculation of required values.

Following implementation of this proposed change, there will be no change in the ability of the EDGs to supply maximum post-accident load demands for 48 hours. The proposed minimum volume of fuel, ≥ 24,000 gallons for two EDGs and ≥ 13,000 gallons for one EDG per train, ensures that a 48-hour supply of fuel is available when the associated standby emergency power source is required to be operable.

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 EDGs and the associated support systems, such as the fuel oil storage and transfer systems, are designed to mitigate accidents and are not accident initiators. Following this change, the EDGs will continue to supply the required maximum post-accident load demand. The current 48-hour fuel supply requirements will be maintained following this change. The new required fuel oil volumes are within the capacities of the fuel oil storage tanks.

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

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

Response: No.

There are two underground fuel oil storage tanks on site. Each tank has a capacity of approximately 35,000 gallons and each common fuel tank supports one EDG train. Start Printed Page 74097Fuel can be manually transferred from one tank to another via a cross-connect valve. Sufficient fuel is maintained between the two tanks to allow one EDG to operate continuously at the required load for seven (7) days. At the proposed minimum required level, which is ≥ 24,000 gallons in the common fuel oil storage tanks for two standby emergency power sources, one tank could provide enough fuel for two EDGs in either Train A or Train B to continue operation for great than 48 hours. At the proposed minimum required level, which is ≥ 13,000 gallons in each fuel oil storage tanks, one tank could provide enough fuel for one EDG in Train A and Train B to continue operation for greater than 48 hours.

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: William Blair, Senior Attorney, NextEra Energy Point Beach, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: Robert J. Pascarelli.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

Date of amendment request: August 27, 2010.

Description of amendment request: The proposed amendment would add a new Action to Technical Specification (TS) 3.7.3, “Control Room Emergency Ventilation (CREV) System,” to permit one or more CREV subsystems to be inoperable for up to 90 days when the inoperability is due to inoperable CREV System High Efficiency Particulate Air (HEPA) filter and/or charcoal absorbers. The proposed TS changes also include an administrative change to correct errors in Unit 2 TS page header information that occurred during issuance of TS pages for a previous amendment.

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.

UFSAR [Updated Finale Safety Analysis Report] Chapter 14, “Plant Safety Analysis,” evaluates operational transients and accidents that result in radiological releases that affect control room occupants. UFSAR section 14.6, “Analysis of Design Basis Accidents—Uprated,” evaluates accidents that release fission products to the environment. The CREV System is not an accident initiator for any of the accidents described. The CREV System processes outside air needed to provide ventilation and pressurization for control room habitability to limit the control room dose during accidents evaluated in the UFSAR. Without crediting the performance of the HEPA filter or charcoal adsorbers, the analyses results concludes that the 30[-]day integrated post-accident doses in the control room are within the limits of 5 rem TEDE [total effective dose equivalent], as specified in 10 CFR 50.67 and GDC [General Design Criterion]-19. The control room dose increase is less than 10 percent; leaving more than 60 percent remaining margin to the regulatory limit.

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 CREV System is a ventilation system that filters outside air used to pressurize the control rooms to provide a protected environment from which operators can control the unit during airborne challenges from radioactivity during accident conditions. The CREV System does not initiate accidents. The proposed amendment allows the CREV HEPA filters and charcoal adsorbers to be repaired or replaced without shutting down the operating unit(s). No new modes of operation are introduced.

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

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

Response: No.

Analyses associated with the prior approval of Alternate Source Term methodology for design basis accident dose consequences previously did not credit the CREV System charcoal adsorbers. Recent analyses have been performed to assess the post-accident 30-day control room dose removing credit for the CREV System HEPA filter. The results indicate a minimal increase in dose consequences (9.5 percent increase) due to removing credit for the CREV System HEPA filter. Even with no credit for either the CREV System HEPA filter or CREV System charcoal filter, the resultant control room dose maintains more than 60 percent margin to the regulatory limit of 5 rem TEDE. As such there is no reduction in a margin of safety for any duration of inoperability of the CREV System HEPA filter or charcoal adsorbers. While the HEPA filter and charcoal adsorbers are not credited for accident mitigation, they remain required by the BFN TS for compliance with the LCO 3.7.3, “Control Room Emergency Ventilation (CREV) System,” further minimizing any potential reduction in a margin of safety.

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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.

NRC Branch Chief: Douglas A. Broaddus.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia

Date of amendment request: October 21, 2010.

Description of amendment request: The proposed action involves the inclusion of the Westinghouse Best-Estimate (BE) Large Break Loss-of-Coolant Accident (LBLOCA) analysis methodology using the Automated Statistical Treatment of Uncertainty Method (ASTRUM) for the analysis of LBLOCA to the list of methodologies approved for reference in the Core Operating Limits Report (COLR) in Technical Specification (TS) 5.6.5.b. This action also removes four obsolete COLR references that supported North Anna Improved Fuel (NAIF) product, Westinghouse Vantage 5, since this product is not planned to be used in future North Anna cores.

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:

[Criterion 1]

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

Response: No.

No physical plant changes are being made as a result of using the Westinghouse Best Estimate Large Break LOCA (BE-LBLOCA) analysis methodology. The proposed TS change simply involves updating the references in TS 5.6.5.b, Core Operating Limits Report (COLR), to reference the Westinghouse BE-LBLOCA analysis methodology, which is an NRC approved methodology, and to delete unnecessary references. Therefore, the probability of LOCA occurrence is not affected by the change. Further, the consequences of a LOCA are not increased, since the BE-LBLOCA analysis has demonstrated that the performance of the Emergency Core Cooling Start Printed Page 74098System (ECCS) continues to conform to the criteria contained in 10 CFR 50.46, “Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors.” No other accident consequence is potentially affected by this change.

Systems will continue to be operated in accordance with current design requirements under the new analysis, therefore no new components or system interactions have been identified that could lead to an increase in the probability of any accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR). No changes were required to the Reactor Protection System (RPS) or Engineering Safety Features (ESF) setpoints because of the new analysis methodology.

An analysis of the LBLOCA accident for North Anna Units 1 and 2 has been performed with the Westinghouse BE-LBLOCA analysis methodology using ASTRUM. The analysis was performed in compliance with the NRC conditions and limitations as identified in WCAP-1 6009-P-A. Based on the analysis results, it is concluded that the North Anna Units 1 and 2 continue to satisfy the limits prescribed by 10 CFR 50.46.

There are no changes to assumptions of the radiological dose calculations. Hence, there is no increase in the predicted radiological consequences of accidents postulated in the UFSAR.

Therefore, neither the probability of occurrence nor the consequences of an accident previously evaluated is significantly increased.

[Criterion 2]

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

Response: No.

The use of the Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not impact any of the applicable design criteria and pertinent licensing basis criteria continue to be met. Demonstrated adherence to the criteria in 10 CFR 50.46 precludes new challenges to components and systems that could introduce a new type of accident. Safety analysis evaluations have demonstrated that the use of Westinghouse BE-LBLOCA analysis methodology with ASTRUM is acceptable. Design and performance criteria continue to be met and no new single failure mechanisms have been created. The use of the Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not involve any alteration to plant equipment or procedures that would introduce any new or unique operational modes or accident precursors. Furthermore, no changes have been made to any RPS or ESF actuation setpoints. Based on this review, it is concluded that no new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes.

Therefore, the possibility for a new or different kind of accident from any accident previously evaluated is not created.

[Criterion 3]

Does this change involve a significant reduction in a margin of safety?

Response: No.

It has been demonstrated that the analytical technique used in the Westinghouse BE-LBLOCA analysis methodology using ASTRUM realistically describes the expected behavior of the reactor system during a postulated LOCA. Uncertainties have been accounted for as required by 10 CFR 50.46. A sufficient number of LOCAs with different break sizes, different locations, and other variations in properties have been considered to provide assurance that the most severe postulated LOCAs have been evaluated. The analysis has demonstrated that the acceptance criteria contained in 10 CFR 50.46 continue to be satisfied.

Therefore, it is concluded that this change does not involve a significant reduction in the margin of safety.

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

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

NRC Branch Chief: Gloria Kulesa.

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, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20854. 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3, Maricopa County, Arizona

Date of application for amendment: April 29, 2010.

Brief description of amendment: The amendments adopted Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) Standard Technical Specification change traveler TSTF-491, Revision 2, “Removal of Main Steam and Main Feedwater Valve Isolation Times from Technical Specifications.” The isolation times will be located outside of the TSs in a document subject to control by the 10 CFR 50.59 process.

Date of issuance: November 5, 2010.

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

Amendment No.: Unit 1—181; Unit 2—181; Unit 3—181.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendment revised the Operating Licenses and Technical Specifications.

Date of initial notice in the Federal Register: July 27, 2010 (75 FR 44024).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 5, 2010.

No significant hazards consideration comments received: No.Start Printed Page 74099

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

Date of application for amendment: June 23, 2010.

Brief description of amendment: Current Technical Specification (TS) 6.5.8, “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 amendment deleted the references to Section XI of the Code and incorporated references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). The amendment also indicates that there may be some nonstandard frequencies utilized in the IST Program in which the provisions of Surveillance Requirement (SR) 3.0.2 are applicable. The changes are consistent with Technical Specification Task Force (TSTF) Technical Change Travelers TSTF-479-A, “Changes to Reflect Revision of 10 CFR 50.55a,” and TSTF-497-A, “Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less.”

Date of issuance: November 5, 2010.

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

Amendment No.: 291.

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

Date of initial notice in the Federal Register: August 10, 2010 (75 FR 48375).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 5, 2010.

No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: May 27, 2010, as supplemented by letter dated August 26, 2010.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.8.3, “Diesel Fuel Oil, Lube Oil, and Starting Air,” by relocating the current stored diesel fuel oil and lube oil numerical volume and level requirements from the TSs to the TS Bases so that it may be modified under licensee control. The TSs have been modified so that the stored diesel fuel oil and lube oil inventory will require that a 7-day supply be available for each diesel generator. Condition A and Condition B in the Action table and Surveillance Requirements (SRs) 3.8.3.1 and 3.8.3.2 are also revised to reflect the above change. The changes are consistent with NRC-approved Revision 1 to Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler TSTF-501, “Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control.” The availability of the TS improvement was announced in the Federal Register on May 26, 2010, as part of the consolidated line item improvement process.

Date of issuance: November 4, 2010.

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

Amendment Nos.: Unit 1—153; Unit 2—153.

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

Date of initial notice in the Federal Register: August 10, 2010 (75 FR 48376). The supplemental letter dated August 26, 2010, 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 amendments is contained in a Safety Evaluation dated November 4, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 17, 2009.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) for Limiting Condition for Operations 3.1.2 “Reactivity Anomalies” changing Surveillance Requirement 3.1.2.1 methodology.

Date of issuance: November 4, 2010.

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

Amendment Nos.: 263 and 207.

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

Date of initial notice in Federal Register: February 23, 2010.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 4, 2010.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 19th day of November 2010.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

End Signature End Preamble

[FR Doc. 2010-29941 Filed 11-29-10; 8:45 am]

BILLING CODE 7590-01-P