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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 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 December 30, 2010 to January 12, 2011. The last biweekly notice was published on January 11, 2011 (76 FR 1644).

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 Start Printed Page 4382publication 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, Public Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852-2738.

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 20852-2738. 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 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://Start Printed Page 4383www.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 EIE, 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 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 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 20852-0238, 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, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852-2738. 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.

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

Date of amendment request: May 20, 2010.

Description of amendment request: The amendments would revise the Technical Specifications (TSs) to allow the reactor building pressure boundary to be opened under administrative controls.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (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 proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to TS 3.6.10 and TS 3.6.16 have no effect upon accident probabilities or consequences. The changes proposed herein will have no impact upon the Reactor Building or AVS [Annulus Ventilation System] relative to the performance of their design functions. These structures/systems will continue to be available and will function as designed during and following all accidents for which their performance is credited in the plant safety analyses. The proposed administrative controls for TS 3.6.16 will ensure the restoration of the Reactor Building pressure Start Printed Page 4384boundary when required, thereby enhancing nuclear safety. No design changes are being made to the plant itself; therefore, there will be no impact upon the probability of any accident occurring. Since the performance of these systems will not be adversely impacted, there will be no impact upon accident consequences.

Criterion 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 to TS 3.6.10 and TS 3.6.16 do not introduce any changes or mechanisms that create the possibility of a new or different kind of accident. No design changes are being made to the plant which would result in the introduction of new accident causal mechanisms. The proposed changes do not introduce any new equipment, any change to existing equipment, or any change to the manner in which the plant is operated. No new effects or malfunctions will therefore be created.

Criterion 3:

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

Response: No.

The proposed changes to TS 3.6.10 and TS 3.6.16 maintain the required design margins of the Reactor Building and AVS for all accidents for which their function is assumed. All required General Design Criteria (GDCs) contained in 10 CFR 50, Appendix A, “General Design Criteria for Nuclear Power Plants” will continue to be satisfied following NRC approval of these proposed changes. In addition, margin of safety is related to the confidence in the fission product barriers to function as designed during and following an accident. These barriers include the fuel cladding, the Reactor Coolant System, and the Containment System. The changes proposed in this submittal have no adverse impact upon the performance of any of these barriers to perform their design functions during or following an accident.

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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Gloria Kulesa.

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

Date of amendment request: September 16, 2010.

Description of amendment request: The amendments would revise Technical Specification 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” to replace the references to the outdated logic per train per doghouse with updated references which reflect the license amendment granted by the U.S. Nuclear Regulatory Commission staff on April 2, 2009.

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

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

Response: No.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configurations of the facility. The proposed changes do not alter or prevent the ability of structures, systems and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. In review of the discussion above (Section 4.1 Significant Hazards Consideration) it can be concluded the probability or consequences of any accident previously evaluated are not increased. This LAR requests administrative changes only.

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

Response: No.

This revision will not impact the accident analysis. The proposed changes will not alter the requirements of the ESFAS or its function during accident conditions. No new or different accidents result from the changes proposed. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or any changes in methods governing normal plant operation. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analyses assumptions. In review of the discussion above (Section 4.1 Significant Hazards Consideration) it can be concluded that these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. This LAR requests administrative changes only.

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

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. In review of the discussion above (Section 4.1 Significant Hazards Consideration) it can be concluded that the proposed changes do not involve a significant reduction in the margin of safety. This LAR requests administrative changes only.

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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Gloria Kulesa.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-369, 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina; 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: June 29, 2010.

Description of amendment request: The amendments would revise Technical Specification (TS) 3.3.1, “Reactor Trip System (RTS) Instrumentation” and TS 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation.”

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 specific Technical Specification changes are associated with (1) the specific Allowable Values for various RTS and ESFAS channels, including instrumentation associated with neutron flux, containment pressure, pressurizer pressure, pressurizer water level, reactor coolant flow, reactor coolant pump underfrequency, steam generator water level, turbine impulse pressure, steam line pressure, and reactor coolant temperature; (2) the addition of specific requirements to be taken if an instrument channel setpoint is outside its predefined as-found tolerance; and (3) the addition of specific requirements regarding resetting of an instrument channel setpoint within an as-left tolerance.

The RTS and ESFAS instrumentation is accident mitigation equipment and does not affect the probability of any accident being initiated. In addition, none of the abovementioned proposed Technical Start Printed Page 4385Specification changes affect the probability of any accident being initiated.

The proposed changes to TS Allowable Values are based on methodology that is consistent with the intent of ISA [Instrument Society of America] Standard RP67.04-1994, Part II, “Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation,” and will preserve assumptions in the applicable accident analyses. None of the proposed changes alter any assumption previously made in the radiological consequences evaluations, nor do they affect mitigation of the radiological consequences of an accident previously evaluated.

In summary, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of any of the proposed changes. The RTS and ESFAS are not capable by itself of initiating any accident. No physical changes to the overall plant are being proposed. No changes to the overall manner in which the plant is operated are being proposed. The proposed changes do not introduce any new failure modes.

Therefore, none of the proposed changes will 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.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions. These barriers include the fuel cladding, the reactor coolant system pressure boundary, and the containment barriers. The proposed changes will not have any impact on these barriers. Plant actuation features and Nominal Trip Setpoints will be unchanged and will actuate prior to exceeding any analytical limits. No accident mitigating equipment will be adversely impacted.

Therefore, existing safety margins will be preserved. None of the proposed changes will 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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Gloria Kulesa.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

Date of amendment request: October 28, 2010.

Description of amendment request: The proposed amendment would modify Clinton Power Station Technical Specifications (TS) Section 3.8.1, “AC Sources Operating,” by revising certain Surveillance Requirements (SR) related to the Division 3 alternating current (AC) Sources. The Division 3 AC Sources are independent sources of offsite and onsite AC power primarily dedicated to the High-Pressure Core Spray (HPCS) system. The TS currently prohibit performing the testing required by SR 3.8.1.8 and SR 3.8.1.12 in Modes 1 or 2, and prohibit performing the testing required by SR 3.8.1.11, SR 3.8.1.16, and SR 3.8.1.19 in Modes 1, 2, or 3. The proposed amendment would remove these Mode restrictions and allow all five of the identified SRs to be performed in any operating Mode for the Division 3 AC Sources.

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: EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of Amendment,” as discussed below:

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

Response: No.

The Division 3 (i.e., HPCS) diesel generator (DG) and its associated emergency loads are accident mitigating features, not accident initiators. Therefore, the proposed TS changes to allow the performance of certain Division 3 AC Sources surveillance testing in any plant operating Mode will not significantly impact the probability of any previously evaluated accident.

The design of plant equipment is not being modified by the proposed changes. As such, the ability of the Division 3 AC Sources to respond to a design basis accident will not be adversely impacted by the proposed changes. Testing procedures include steps to ensure that injection into the reactor vessel is precluded. The proposed changes to the TS surveillance testing requirements for the Division 3 AC Sources do not affect the operability requirements for the AC Sources, as verification of such operability will continue to be performed as required. Continued verification of operability supports the capability of the Division 3 AC Sources to perform their required functions of providing emergency power to HPCS system equipment, consistent with the plant safety analyses. Limiting testing to only one AC Source at a time ensures that design basis requirements are met. Should a fault occur while testing the Division 3 AC Sources, there would be no significant impact on any accident consequences since the other two divisional AC Sources and associated emergency loads would be available to provide the minimum safety functions necessary to shut down the unit and maintain it in a safe shutdown condition.

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

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

Response: No.

No changes are being made to the plant that would introduce any new accident causal mechanisms. Equipment will be operated in the same configuration with the exception of the plant operating mode in which the Division 3 AC Sources surveillance testing is conducted. Performance of these surveillances tests while online will continue to verify operability of the Division 3 AC Sources. The proposed amendment does not impact any plant systems that are accident initiators and does not adversely impact any accident mitigating systems.

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

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

Response: No.

Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed changes to the TS surveillance testing requirements for the Division 3 AC Sources do not affect the operability requirements for the AC Sources, as verification of such operability will continue to be performed as required. Continued verification of operability supports the capability of the Division 3 AC Sources to perform their required function of providing emergency power to HPCS system equipment, consistent with the plant safety analyses. Consequently, the performance of the fission product barriers will not be adversely impacted by implementation of the proposed amendment. In addition, the proposed changes do not alter setpoints or limits established or assumed by the accident analysis. Further, performing Division 3 AC Sources surveillance activities online increases the Division 3 DG and HPCS system availability during refueling outages and allows the testing of the Division 3 systems to be conducted when both Division 1 and 2 systems are required to be OPERABLE.

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

Start Printed Page 4386

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. Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: July 16, 2010.

Description of amendment request: The amendments would revise the Technical Specifications (TSs) to adopt Nuclear Regulatory Commission (NRC)-approved Revision 3 to Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF-448, “Control Room Envelope Habitability.” The proposed amendments include changes to the TS requirements related to control room envelope (CRE) habitability in TS 3/4.7.5, “Control Room Emergency Ventilation System (CREVS),” and TS Section 6.8, “Administrative Controls—Procedures and Programs.” This submittal satisfies the commitment identified in FPL's letter dated August 10, 2007, to adopt the applicable portions of TSTF-448. Additionally, this application updates the original submittal of license amendment request 194 dated September 26, 2008, in response to an NRC request for additional information to remove any reference of unapproved TSTF-508, which has been done.

The NRC staff published a notice of opportunity for comment in the Federal Register on October 17, 2006 (71 FR 61075), on possible amendments adopting TSTF-448, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on January 17, 2007 (72 FR 2022). The licensee affirmed the applicability of the following NSHC determination in its application dated July 16, 2010.

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

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

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not 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.

The proposed change does not impact the accident analysis. The proposed change does not alter the required mitigation capability of the CRE emergency ventilation system, or its functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice.

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

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

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

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

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

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: October 29, 2010.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.8.4, “DC [Direct Current] Sources—Operating,” and TS 3.8.6, “Battery Cell Parameters.” Specifically, the proposed changes would replace non-conservative minimum voltages in Surveillance Requirement 3.8.4.1 for the 125 volt direct current (V DC) and 250 V DC essential batteries, and the non-conservative battery specific gravity values listed in TS Table 3.8.6-1, “Battery Cell Parameter Requirements.”

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

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

Response: No.

Performing surveillances that verify terminal voltage and specific gravity of batteries is not a precursor of any accident previously evaluated. Restoring battery limits to conservative values does not significantly affect the method of performing the surveillances, such that the probability of an accident would be affected. Therefore, the proposed changes do not result in a significant increase in the probability of an accident previously evaluated.

Restoring battery limits to conservative values so that batteries are maintained in Start Printed Page 4387accordance with plant design basis ensures they provide the power assumed in design basis accident mitigation calculations. Therefore, the change does not involve a significant increase in the consequences of an accident previously evaluated.

NPPD [Nebraska Public Power District] concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve any modification to the plant or equipment or how they are operated. Therefore, NPPD concludes that these proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change will continue to ensure station batteries are able to perform their design function as assumed in calculations that evaluate their function during design basis accidents. The proposed change actually increases the margin of safety by restoring conservatisms inherent in battery design and manufacturer's recommendations. Based on this, the ability of CNS [Cooper Nuclear Station] to mitigate the design basis accidents that rely on operation of the station batteries is not adversely impacted. Therefore, NPPD concludes that these proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: July 12, 2010.

Description of amendment request: This notice is being reissued in its entirety due to missing statements from the description of the amendment request in the notice published in the Federal Register on December 28, 2010 (75 FR 81671). The proposed amendment would modify Item 1 of Table 2-5, “Instrumentation Operating Requirements for Other Safety Feature Functions,” of Technical Specification (TS) 2.15, “Instrumentation and Control Systems,” to provide new Note (e), and Surveillance Requirement (SR) Items 1 and 2 of Table 3-3, “Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls,” of TS 3.1, “Instrumentation and Control,” which pertain to operability of the primary and secondary control element assembly (CEA) position indication system (CEAPIS) channels. A new SR is proposed for Item 4 of Table 3-3 of TS 3.1, which will verify the position of CEAs each shift. The proposed amendment will ensure that CEA alignment is maintained during power operations so that the power distribution and reactivity limits defined by the design power peaking and shutdown margin (SDM) limits are preserved. The proposed amendment would also revise TS 2.10.2(7)c regarding actions to be taken when the regulating CEA groups are inserted below the Long Term Insertion Limit. The TS would be revised to require actions to be taken when either time interval is exceeded, which would also make TS 2.10.2(7)c more consistent with Combustion Engineering (CE) Standard Technical Specifications (STS).

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

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

Response: No.

The proposed amendment will allow plant operation to continue when a CEAPIS channel is inoperable by requiring prompt verification of CEA positions following CEA movement. CEAs are most likely to become misaligned during movement and therefore, this change will cause CEA alignment errors to be promptly detected and corrected. It is appropriate to clarify that CEAPIS channels are not subject to the requirements of TS 2.15(1), (2), and (3) as they are not designed to be placed in trip or bypass, nor are they engineered safety feature (ESF) or isolation logic subsystems.

The proposed amendment does not alter the requirements of TS 2.15(4) regarding the rod block function of the secondary CEAPIS channel. Should the secondary CEAPIS channel or its rod block function be inoperable, several additional CEA deviation events are possible. However, this situation is already addressed by TS 2.15(4), which requires the CEAs (rods) to be maintained fully withdrawn with the control rod drive system mode switch in the off position except when manual motion of CEA Group 4 is required to control axial power distribution. This is the same position that the CEAs must be in (fully withdrawn) when the plant is at power (Mode 1) in order to utilize distributed control system (DCS) core mimic to CHANNEL CHECK the CEAPIS channels.

If it was not possible to use DCS core mimic to verify the primary CEAPIS channel as would be the case if CEA Group 4 was inserted to control axial power distribution, then the primary CEAPIS channel would be declared inoperable when the CHANNEL CHECK could not be accomplished. The plant would then be placed in hot shutdown (Mode 3) within 12 hours in accordance with TS 2.15(4). Therefore, although the proposed amendment will allow a CEAPIS channel to be inoperable indefinitely, there is no significant increase in the probability or consequences of an accident as the requirements of TS 2.15(4) will continue to be met. This serves to prevent the type of CEA deviation events that the rod block function was designed for.

Replacing the current method of verifying CEAPIS data with the defined term CHANNEL CHECK is an improvement that provides additional flexibility without weakening the intent of the surveillance. As a result, when it is feasible to obtain CEA position indication from DCS core mimic (i.e., when the CEAs are either fully inserted or fully withdrawn), the primary and secondary CEAPIS channels will be compared with DCS core mimic indication as well as each other.

As an additional means of verifying CEA positions, DCS core mimic indication provides added confidence that the CEAs are in the indicated positions. Should the primary or secondary CEAPIS channel become inoperable, the accuracy and reliability of DCS core mimic indication is assured by its previous comparison with both OPERABLE channels. Comparison of the OPERABLE CEAPIS channel with DCS core mimic will satisfy the required CHANNEL CHECK and allow continued operation while the inoperable channel is repaired. The proposed amendment ensures that the CEA alignment required by TS 2.10.2(4) is met each shift by requiring all full length (shutdown and regulating) CEAs to be positioned within 12 inches of all other CEAs in the group.

The change proposed for TS 2.10.2(7)c incorporates more conservative wording to ensure that the regulating CEA groups are maintained within the Long Term Insertion Limit. The proposed change will ensure that corrective actions are taken if either time interval is exceeded and makes TS 2.10.2(7)c more consistent with CE STS.

The proposed amendment does not alter the plant configuration, require new plant equipment to be installed, alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected.

As an additional means of verifying primary and secondary CEAPIS data, DCS core mimic indication increases confidence in the reliability of CEAPIS data.

The proposed amendment will help minimize unplanned shutdowns that can Start Printed Page 4388cause plant transients yet continues to ensure that power distribution and reactivity limits are maintained. Therefore, it is concluded that this 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 does not change the design function or operation of the primary or secondary CEAPIS channels. If one CEAPIS channel should become inoperable, the position of CEAs will be verified within 15 minutes of any CEA movement to quickly detect and correct CEA alignment errors. Data from each CEAPIS channel will continue to be compared to the other channel each shift as before. However, a CHANNEL CHECK will require that CEAPIS channel data also be compared with DCS core mimic indication when it is available. Thus, when the CEAPIS channels are required to be OPERABLE, there will be at least two means of verifying the position of CEAs or else appropriate actions must be taken. The CEA alignment required by TS 2.10.2(4) is assured by requiring verification each shift that all full length (shutdown and regulating) CEAs are positioned within 12 inches of all other CEAs in the group.

No changes are proposed to testing and calibration of the CEAPIS channels and these requirements will continue to ensure that they are capable of performing their design function. Use of the defined term CHANNEL CHECK is an appropriate surveillance method as it requires that the channel be compared with other independent channels measuring the same variable where feasible. DCS core mimic is a diverse, accurate and reliable means of verifying CEA positions when the CEAs are fully inserted or fully withdrawn. The change proposed for TS 2.10.2(7)c ensures that appropriate corrective actions are taken when the regulating CEA groups are below the Long Term Insertion Limit in excess of either of the specified time intervals.

No new structures, systems, or components (SSCs) are being installed, and no credible new failure mechanisms, malfunctions, or accident initiators are created. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

When a CEAPIS channel is inoperable, the proposed amendment allows plant operation to continue but requires more frequent verification of CEA positions following any CEA movement, which is when CEAs are most likely to become misaligned. This will enable CEA alignment errors to be detected and corrected more promptly. As CEAPIS channels are not designed to be placed in trip or bypass, nor are they engineered safety feature (ESF) or isolation logic subsystems, it is appropriate to clarify that TS 2.15(1), (2), and (3) do not apply. FCS normally operates with the CEAs fully withdrawn and maintains reactivity control by adjusting reactor coolant system (RCS) boric acid concentration. When the CEAs are fully withdrawn (or fully inserted), DCS core mimic indication provides accurate and reliable indication of CEA positions suitable for comparison with the primary and secondary CEAPIS channels. Thus, even with one CEAPIS channel inoperable, a diverse means of verifying the accuracy of the OPERABLE CEAPIS channel will be available. The accuracy and reliability of DCS core mimic is assured by testing conducted each refueling outage with continued assurance provided by comparison with primary and secondary CEAPIS each shift.

The change also ensures that the CEA alignment required by TS 2.10.2(4) is met each shift by requiring all full length (shutdown and regulating) CEAs to be positioned within 12 inches of all other CEAs in the group. The proposed amendment does not alter the TS 2.15(4) requirement to place the reactor in hot shutdown in the event that both CEAPIS channels are inoperable. The change proposed for TS 2.10.2(7)c incorporates more conservative wording to ensure that the regulating CEA groups are maintained within the Long Term Insertion Limit.

The proposed amendment will help minimize unplanned shutdowns that can cause plant transients yet continues to ensure that power distribution and reactivity limits are maintained. The proposed amendment does not alter the plant configuration, require new plant equipment to be installed, alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: August 16, 2010, as supplemented by letter dated September 27, 2010.

Description of amendment request: The proposed amendment would remove the Technical Specification (TS) limiting condition for operation (LCO) 2.15, “Instrumentation and Control Systems,” Table 2-5, “Instrumentation Operating Requirements for Other Safety Feature Functions,” Items 3, 4, and 5, the associated Notes a, b, c, and d, and the associated footnote, for power-operated relief valve (PORV) and pressurizer safety valve (PSV) acoustic position indication and tail pipe temperature from the Fort Calhoun Station (FCS) TS. The proposed amendment would also revise the surveillance requirement (SR), TS 3.1, “Instrumentation and Control,” Table 3-3, “Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls,” Items 21, 23, and 24 for PORV Operation and Acoustic Position Indication, Safety Valve Acoustic Position Indication, and PORV/Safety Valve Tail Pipe Temperature, respectively. Specifically, Table 3-3, Item 21 will be revised to reflect the performance of the PORV operation channel functional test on its existing refueling frequency and deletes the monthly frequency denoted in the TS for the acoustic position indication which would also be more aligned with NUREG-1432, “Standard Technical Specifications, Combustion Engineering Plants,” Revision 3, for PORV operation; and Items 21, 23, and 24 will be revised to relocate the acoustic position indication and tail pipe temperature indication SRs from the FCS TS. In conjunction with the proposed TS changes, operability and surveillance requirements for the acoustic position indication and tail pipe temperature indication instrumentation would be incorporated into the FCS Updated Safety Analysis Report (USAR) and associated plant procedures.

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 safety valve acoustic position indication does not affect the operation of its associated spring-loaded safety valve. As such, the proposed change does not increase the probability of an accident. The acoustic monitor and tail pipe temperature indication are only two of the indications used to identify that a safety valve is open. Other indications are available to the operators and alarm in the control room. The acoustic monitor is only one of the indications that the abnormal and emergency procedures direct operators to use to diagnose the opening of a safety valve. The failure of the power operated relief valve (PORV)/safety valve position instrumentation is not assumed to be an initiator of any analyzed event in the Updated Safety Analysis Report (USAR). The proposed changes do not alter Start Printed Page 4389the physical design of the PORVs/safety valves or any other plant structure, system or component (SSC). The changes would remove the PORV/safety valve position indicator operability and surveillance requirements from the Fort Calhoun Station (FCS) Technical Specifications (TS), and incorporate the requirements for this instrumentation into a licensee-controlled document under the control of 10 CFR 50.59.

The proposed changes conform to the Nuclear Regulatory Commission's (NRC's) regulatory guidance regarding the content of plant TS as identified in 10 CFR 50.36 and NRC publication NUREG-1432.

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 alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Hence, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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 instrumentation is not needed for manual operator actions necessary for safety systems to accomplish their safety function for the design basis accident events. The acoustic position indicator and tail-pipe temperature instrumentation provides only alarm and PORV/safety valve position indication, and does not provide an input to any automatic trip function. Diverse means are available to monitor PORV/safety valve position, and operability and surveillance requirements will be established in a licensee-controlled document to ensure the reliability of the PORV/safety valve position monitoring capability. Changes to these requirements will be subject to the controls of 10 CFR 50.59, providing the appropriate level of regulatory control. In addition, the PORV operation is currently tested on a refueling frequency, which is aligned with the surveillance requirements provided in NUREG-1432.

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.

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

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

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

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

Date of amendment request: January 27, 2010.

Brief description of amendment request: The amendment revises Section 2.E. of the Palisades Nuclear Plant (PNP) Renewed Facility Operating License to remove the name of the former operator of the plant in the title of the PNP physical security plan and replace it with Entergy Nuclear. The change also removes the security plan revision number and the date the plan was submitted to the Nuclear Regulatory Commission.

Date of publication of individual notice in Federal Register: November 18, 2010 (75 FR 70708).

Expiration date of individual: January 17, 2011

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.

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

Date of application of amendments: May 30, 2008, as supplemented by letters dated October 31, 2008, January 30, 2009, February 9, 2009, February 23, 2009, May 31, 2009, August 3, 2009, September 29, 2009, and November 30, 2009. By letter dated April 14, 2010, the licensee resubmitted the application and superseded the contents of the application submitted by letter dated May 30, 2008, as supplemented October 31, 2008. This resubmitted application, however, does not supersede the supplements dated January 30, 2009, February 9, 2009, February 23, 2009, May 31, 2009, August 3, 2009, Start Printed Page 4390September 29, 2009, and November 30, 2009. By letters dated September 13, 2010, September 27, 2010, October 14, 2010, November 19, 2010, and December 22, 2010, the licensee supplemented the April 14, 2010 application.

Brief description of amendments: The amendments revised the licenses and Technical Specifications to allow the licensee to maintain a fire protection program in accordance with 10 CFR 50.48(c) for the Oconee Nuclear Station, Units 1, 2, and 3.

Date of Issuance: December 29, 2010.

Effective date: As of the date of issuance and shall be fully implemented prior to January 1, 2013.

Amendment Nos.: Unit 1—371, Unit 2—373, Unit 3—372.

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

Date of initial notice in Federal Register: October 28, 2010 (75 FR 66395).

The supplements dated September 13, 2010, September 27, 2010, October 14, 2010, November 19, 2010, and December 22, 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.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 29, 2010.

No significant hazards consideration comments received: No.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: July 22, 2010.

Brief description of amendment: The amendment revised Limiting Condition for Operation (LCO) 3.10.1, “Inservice Leak and Hydrostatic Testing Operation,” and the associated Bases, to expand its scope to include provisions for temperature excursions greater than 200 degrees Fahrenheit as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4. The change is consistent with NRC-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-484, “Use of TS 3.10.1 for Scram Time Testing Activities,” that was announced in the Federal Register on October 27, 2006 (71 FR 63050), as part of the Consolidated Line Item Improvement Process (CLIIP).

Date of issuance: January 5, 2011.

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

Amendment No.: 170.

Facility Operating License No. NPF-47: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: October 5, 2010 (75 FR 61524).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 5, 2011.

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: March 24, 2010, supplemented by letters dated July 29, 2010, and September 27, 2010.

Brief description of amendment: The changes revise the TMI-1 technical specifications to relocate certain surveillance frequencies to a licensee-controlled program through the implementation of Nuclear Energy Institute 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.” The changes are consistent with U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications change TSTF-425, “Relocate Surveillance Frequencies to Licensee Control—Risk Informed Technical Specifications Task Force Initiative 5b,” Revision 3.

Date of issuance: January 12, 2011.

Effective date: Immediately, and shall be implemented within 120 days.

Amendment No.: 274.

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

Date of initial notice in Federal Register: May 18, 2010 (75 FR 27829).

The supplements dated July 29, 2010, and September 27, 2010, 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 determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 12, 2011.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 13th day of January 2011.

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. 2011-1480 Filed 1-24-11; 8:45 am]

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