Skip to Content

Notice

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

Document Details

Information about this document as published in the Federal Register.

Published Document

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

Start Preamble

I. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any Start Printed Page 61522amendments 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 September 9, 2010, to September 22, 2010. The last biweekly notice was published on September 21, 2010 (75 FR57521).

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

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

Written comments may be submitted by mail to the Chief, 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 File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the 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 Start Printed Page 61523consideration, 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 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 (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 Start Printed Page 61524Room 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, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: May 28, 2010.

Description of amendment request: The amendments would revise the Technical Specifications (TS) to allow manual operation of the containment spray system (CSS) and to change the setpoints for the refueling water storage tank (RWST).

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

Response: No.

The CSS and RWST are accident mitigation equipment. As such, changes in operation of these systems cannot have an impact on the probability of an accident.

The RWST will continue to comply with all applicable regulatory requirements and design criteria following approval of the proposed changes (e.g., train separation, redundancy, and single failure). The water level on the containment floor will be higher at the start of transfer to the containment sump but will remain below the maximum design level analyzed for equipment submergence. The change in the sump pH will not result in a significant increase in radiological consequences of a LOCA [loss-of-coolant accident]. Therefore, the design functions performed by the equipment are not changed.

The proposed change alters the method of controlling the safety system following a design basis event so that manual actions are substituted for automatic actions. Calculations and simulator exercises confirm these actions will be taken within the appropriate scenario sequence timing to provide containment cooling and source term reduction.

The delay in CS [containment spray] operation will result in an increase in containment temperature, containment pressure, offsite dose, and control room dose during a LOCA or high energy line break inside containment. Containment analyses have been performed to demonstrate that containment pressure and temperature remain within the design limits and there is no significant impact on the environmental qualification for equipment inside containment. The reduction in fission product removal due to delayed CS operation does not result in exceeding the offsite dose and control room dose limits in 10 CFR 50.67. The analysis of the change in containment conditions due to a single failure of an operating spray pump and the suspension of CS determined that the pressure remained below the design limits.

The proposed change to adopt [Technical Specification Task Force] TSTF-493, Rev. 4, on a limited basis clarifies requirements for instrumentation to ensure the instrumentation will actuate as assumed in the safety analysis. Instruments are not an assumed initiator of any accident previously evaluated. As a result, the proposed change will not increase the probability of an accident previously evaluated. The proposed change will ensure that the instruments actuate as assumed to mitigate the accidents previously evaluated. As a result, the proposed change will not increase the consequences of an accident previously evaluated.

Based on this discussion, the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated.

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 modification to the low level setpoint will not install any new plant equipment. The setpoint will continue to be included within the engineered safeguards features instrumentation and monitored according to the applicable surveillance requirements. The evaluation of the new level setpoint and the change in the switchover sequence concluded that the equipment aligned to the sump will continue to have sufficient suction pressure prior to containment sump suction switchover. The design of the RWST low level instrumentation complies with all applicable regulatory requirements and design criteria.

The overall function of the CSS is not changed by this proposed amendment. The proposed change alters the method of controlling the safety system following a design basis event so that manual actions are substituted for automatic actions. Calculations confirm that these actions will be taken within the appropriate scenario sequence timing to provide containment cooling and source term reduction with no significant increase in radiological consequences and without exceeding containment design limits.

The proposed change to adopt TSTF-493, Rev. 4 on a limited basis does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change does not alter assumptions made in the safety analysis but ensures that the instruments behave as assumed in the accident analysis. The proposed change is consistent with the safety analysis assumptions.

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

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

Response: No.

The proposed change will increase the calculated radiological dose at the site boundary and in the control room. However, the calculations demonstrate that the dose consequences at the site boundary, low population zone, and control room remain within regulatory acceptance limits of 10 CFR 50.67.

Additional analysis concluded:

  • Peak containment pressure for analyzed design basis accidents will not be significantly increased and containment design limits will not be exceeded.
  • Assumptions used in the environmental qualification of equipment exposed to the containment atmosphere remain bounding.
  • Pumps aligned to the RWST and to the containment sump will have adequate suction pressure.
  • The CSS will retain its ability to undergo all appropriate testing requirements following implementation of the proposed amendment. These testing requirements are conducted in accordance with the McGuire Inservice Testing Program and TS 3.6.6.

It is estimated that the implementation of this license amendment request will result in an approximate 22% reduction in core damage frequency. This amendment request is based on the Nuclear Energy Institute (NEI) and the Pressurized Water Reactor (PWR) Owners Group initiative to extend the post-Loss of Coolant Accident (LOCA) injection phase and delay the onset of the containment sump recirculation phase.

The proposed change to adopt TSTF-493, Rev. 4 on a limited basis clarifies the requirements for instrumentation to ensure the instrumentation will actuate as assumed in the accident analysis. No change is made to the accident analysis assumptions and no margin of safety is reduced as part of this 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, 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.

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.

Description of amendment request: The proposed amendment would revise Start Printed Page 61525Limiting 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 (°F) 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 proposed change is consistent with NRC-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specification Traveller, TSTF-484, “Use of TS 3.10.1 for Scram Time Testing Activities,” that was announced in the Federal Register on October 27, 2001 (71 FR 63050), as part of the consolidated Line Item Improvement Process (CCIIP).

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

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

Response: No.

Technical Specifications currently allow for operation at > 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or consequences of an accident previously evaluated.

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

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

Response: No.

Technical Specifications currently allow for operation at > 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. The extended allowances would result from operations that commence at reduced temperatures, but approach the normal MODE 4 limit of 200 °F prior to completion of the inspections or testing. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

Technical Specifications currently allow for operation at > 200 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an inservice leak or hydrostatic test prior to power operation, results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure.

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: Joseph A. Aluise, Assistant General Counsel—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.

NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: August 19, 2010.

Description of amendment request: The proposed amendment would revise Technical Specifications to be consistent with Standard Technical Specifications 3.6.1.8 “Suppression Chamber-to-Drywell Vacuum Breakers” and 3.6.2.5 “Drywell-to-Suppression Chamber Differential Pressure,” along with the associated Bases, of NUREG-1433, Revision 3, “Standard Technical Specifications General Electric Plants, BWR/4,” modified to account for plant specific design details.

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 does not significantly increase the probability or consequences of an accident since it does not involve a modification to any plant equipment or affect how plant systems or components are operated. No design functions or design parameters are affected by the proposed amendment. The proposed amendment involves the operation and testing of Primary Containment systems but does not impact containment design or performance requirements. 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.

The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. No new or different types of equipment will be installed and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. The proposed amendment involves the operation and testing of Primary Containment systems but does not alter the way that the systems are operated or how the tests are performed. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change ensures that the safety functions of the pressure suppression chamber-drywell vacuum breakers and drywell-suppression chamber differential pressure are fulfilled by incorporating the guidance of NUREG-1433. The proposed amendment does not involve a physical modification of the plant and does not change the design or function of any component or system. Therefore, the proposed amendment will not involve a significant reduction in the margin of safety.

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

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Nancy Salgado.Start Printed Page 61526

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

Date of amendment request: August 10, 2010.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.9.3, “Reactor Building Penetrations,” to allow reactor building flow path(s) providing direct access from the reactor building atmosphere to the outside atmosphere to be unisolated under administrative control, during movement of irradiated fuel assemblies. The proposed change is consistent with Technical Specification Task Force (TSTF) Technical Change Traveler 312, Revision 1, “Administratively Control Containment Penetrations.”

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

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

Response: No.

The status of the penetration flow paths during fuel movement in the reactor building has no affect on the probability of the occurrence of any accident previously evaluated. The proposed change does not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. Since the consequences of a fuel handling accident (FHA) inside the reactor building with open penetrations flow paths is bounded by the current FHA analyses and the probability of an accident is not affected by the status of the penetration flow paths, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The open reactor building penetration flow paths are not accident initiators. The proposed allowance to open the reactor building penetrations during fuel movement inside the reactor building will not adversely affect plant safety functions or equipment operating practices such that a new or different accident could be created. Therefore, the proposed change does not create the possibility of an accident of a different kind than previously evaluated.

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

Response: No.

Technical Specification (TS) 3.9.3 closure requirements for reactor building penetrations ensure that the consequences of a postulated FHA inside the reactor building during irradiated fuel handling activities are minimized. The Limiting Condition for Operation establishes reactor building closure requirements, which limit the potential escape paths for fission products by ensuring that there is at least one integral barrier to the release of radioactive material. The proposed change to allow the reactor building penetration flow paths to be open during refueling operations under administrative controls does not significantly affect the expected dose consequences of a FHA because the limiting FHA does not credit reactor building closure or filtration. The proposed administrative controls provide assurance that prompt closure of the penetration flow paths will be accomplished in the event of a[n] FHA inside the reactor building. The provisions to promptly isolate open penetration flow paths provide assurance that the offsite dose consequences of a[n] FHA inside containment will be minimized. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Joseph A. Aluise, Assistant General Counsel—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: June 29, 2010, as supplemented on August 24, 2010.

Description of amendment request: The proposed amendments would revise Technical Specifications (TS) Section 3.4.12, “Low Temperature Overpressure Protection (LTOP) System,” to correct an inconsistency between the TS, and implementation of procedures and administrative controls for Safety Injection (SI) pumps required to mitigate a postulated loss of decay heat removal during mid-loop operation as discussed in NRC Generic Letter (GL) 88-17, “Loss of Decay Heat Removal.”

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

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

Response: No.

The proposed change does not result in any physical changes to safety related structures, systems, or components. The proposed change revises TS 3.4.12 to correct an inconsistency between the TS, and implementation of procedures and administrative controls for SI pumps required to mitigate a postulated loss of decay heat removal during mid-loop operation as discussed in GL 88-17. Specifically, the proposed change adds a note to TS LCO [limiting condition for operation] 3.4.12 that states: “For the purpose of protecting the decay heat removal function, one or more SI pumps may be capable of injecting into the RCS in MODE 5 and MODE 6 when the reactor vessel head is on provided pressurizer level is ≤ 5 percent.” The proposed change corrects an oversight introduced during the conversion of the Braidwood Station and Byron Station TS to the ITS [Improved TS].

The probability of occurrence of an accident is not increased since the proposed change will continue to require that no SI pumps are capable of injecting into the RCS in Modes 5 and 6 with pressurizer level greater than 5 percent.

The NRC has previously evaluated the allowance for one or more SI pumps to be capable of injecting into the RCS in Mode 5 or Mode 6 when the reactor vessel head is on provided pressurizer level is ≤ 5 percent for the Braidwood Station and Byron Station. In a safety evaluation dated August 31, 1990, related to Braidwood Station, Units 1 and 2, Amendment 25, and Byron Station, Units 1 and 2, Amendment 38, the NRC concluded that allowing SI pump capability to inject into the RCS in Mode 5 or Mode 6 when the reactor vessel head is on provided pressurizer level is ≤ 5 percent was acceptable. The availability of SI pumps under these circumstances does not present a concern regarding cold overpressure protection since sufficient air volume exists which allows Operations personnel time to mitigate the transient. This is in contrast to the analyzed cold overpressure transients, in which the RCS is assumed to be water solid at the onset of the event.

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

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

Response: No.

The proposed change revises TS 3.4.12 to correct an inconsistency between the TS, and implementation of procedures and administrative controls for SI pumps required to mitigate a postulated loss of decay heat removal during mid-loop operation as discussed in GL 88-17. Specifically, the proposed change adds a note to TS LCO 3.4.12 that states: “For the purpose of protecting the decay heat removal function, one or more SI pumps may be Start Printed Page 61527capable of injecting into the RCS in MODE 5 and MODE 6 when the reactor vessel head is on provided pressurizer level is ≤ 5 percent.” The proposed change corrects an oversight introduced during the conversion of the Braidwood Station and Byron Station TS to the ITS.

The proposed change is necessary for the purpose of mitigating the consequences of a loss of decay heat removal during mid-loop operations. Operation of at least one SI pump is required in some cases to prevent the core from uncovering. The only new configuration allowed by the proposed change is the potential of having one or more SI pumps available in Modes 5 and 6 with pressurizer level ≤ 5 percent. The potential overpressurization accident has been analyzed and accounted for by requiring pressurizer level to be ≤ 5 percent if one or more SI pumps are available.

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

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

Response: No.

The proposed change revises TS 3.4.12 to correct an inconsistency between the TS, and implementation of procedures and administrative controls for SI pumps required to mitigate a postulated loss of decay heat removal during mid-loop operation as discussed in GL 88-17. Specifically, the proposed change adds a note to TS LCO 3.4.12 that states: “For the purpose of protecting the decay heat removal function, one or more SI pumps may be capable of injecting into the RCS in MODE 5 and MODE 6 when the reactor vessel head is on provided pressurizer level is ≤ 5 percent.” The proposed change corrects an oversight introduced during the conversion of the Braidwood Station and Byron Station TS to the ITS.

The proposed note allows one or more SI pumps to be capable of injecting into the RCS only when pressurizer level is ≤ 5 percent in Mode 5 and Mode 6 when the reactor vessel head is on. This provides protection to limit coolant input capacity during shutdown in which a pressure fluctuation due to coolant input from the SI pumps could occur more quickly than an operator could react, while providing an allowance for one or more SI pumps to be capable of injecting into the RCS during conditions in which a loss of decay heat removal could result in rapid core uncovery.

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

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

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

NRC Branch Chief: 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: August 5, 2010.

Description of amendment request: The proposed amendments would revise technical specification (TS) 5.5.1 Fuel Storage—Criticality, to include new spent fuel storage patterns that account for both the increase in fuel maximum enrichment from 4.5 weight percentage (wt%) U-235 to 5.0 wt% U-235 and the impact on the fuel of higher power operation proposed under the Extended Power Uprate (EPU) project. Although the fuel storage has been analyzed at the higher fuel enrichment in the new criticality analysis, the fuel enrichment limit of 4.5 wt% U-235 specified in TS 5.5.1 will not be changed under this license amendment request. The proposed TS changes and a new supporting criticality analysis are being submitted to revise the current licensing basis analysis for both new fuel and spent fuel pool storage.

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

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

No. The proposed amendments do not change or modify the fuel, fuel handling processes, fuel storage racks, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the spent fuel pool cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously evaluated events and accidents:

a. A fuel handling accident (FHA),

b. A cask drop accident,

c. A fuel mispositioning event,

d. A spent fuel pool boron dilution event,

e. A seismic event, and

f. A loss of spent fuel pool cooling event.

Although the proposed amendment will require increased handling of the fuel, the probability of a FHA is not significantly increased because the implementation of the proposed amendment will employ the same equipment and process to handle fuel assemblies that is currently used. Also, tests have confirmed that the Metamic inserts can be installed and removed without damaging the host fuel assemblies. The FHA radiological dose consequences associated with fuel enrichment at this level were addressed in LAR [license amendment request] 196 on Alternative Source Term implementation at EPU conditions and remain unchanged. Therefore, the proposed amendments do not significantly increase the probability or consequences of a FHA.

The proposed amendments do not increase the probability of dropping a fuel transfer cask because they do not introduce any new heavy loads to the SFP and do not affect heavy load handling processes. Also, the insertion of Metamic rack inserts does not increase the consequences of the cask drop accident because the radiological source term of that accident is developed from a non-mechanistically derived quantity of damaged fuel stored in the spent fuel pool. Therefore, the proposed amendments do not significantly increase the probability or consequences of a cask drop accident.

Operation in accordance with the proposed amendment will not change the probability of a fuel mispositioning event because fuel movement will continue to be controlled by approved fuel handling procedures. These procedures continue to require identification of the initial and target locations for each fuel assembly that is moved. The consequences of a fuel mispositioning event are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for the worst-case fuel mispositioning event.

Operation in accordance with the proposed amendment will not change the probability of a boron dilution event because the systems and events that could affect spent fuel pool soluble boron are unchanged. The consequences of a boron dilution event are unchanged because the proposed amendment reduces the soluble boron requirement below the currently required value and the maximum possible water volume displaced by the inserts is an insignificant fraction of the total spent fuel pool water volume.

Operation in accordance with the proposed amendment will not change the probability of a seismic event. The consequences of a seismic event are not significantly increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks with Metamic inserts is not appreciably increased. Seismic analyses demonstrate adequate stress levels in the storage racks when inserts are installed.

Operation in accordance with the proposed amendment will not change the probability of a loss of SFP cooling event because the systems and events that could affect SFP cooling are unchanged. The consequences are not significantly increased because there are no changes in the SFP heat load or SFP cooling systems, structures or components. Furthermore, conservative analyses indicate that the current design requirements and criteria continue to be met with the Metamic inserts installed.

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

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

No. The proposed amendments do not change or modify the fuel, fuel handling processes, fuel racks, number of fuel assemblies that may be stored in the pool, Start Printed Page 61528decay heat generation rate, or the spent fuel pool cooling and cleanup system. The effects of operating with the proposed amendment are listed below. The proposed amendments were evaluated for the potential of each effect to create the possibility of a new or different kind of accident:

a. Addition of inserts to the fuel storage racks,

b. New storage patterns,

c. Additional weight from the inserts,

d. Insert movement above fuel, and

e. Displacement of fuel pool water by the inserts.

Each insert will be placed between a fuel assembly and the storage cell wall, taking up some of the space available on two sides of the fuel assembly. Tests confirm that the insert can be installed and removed without damaging the fuel assembly. Analyses demonstrate that the presence of the inserts does not adversely affect spent fuel cooling, seismic capability, or subcriticality. The aluminum (alloy 6061) and boron carbide materials of construction have been shown to be compatible with nuclear fuel, storage racks and spent fuel pool environments, and generate no adverse material interactions. Therefore, placing the inserts into the spent fuel pool storage racks cannot cause a new or different kind of accident.

Operation with the proposed fuel storage patterns will not create a new or different kind of accident because fuel movement will continue to be controlled by approved fuel handling procedures. These procedures continue to require identification of the initial and target locations for each fuel assembly that is moved. There are no changes in the criteria or design requirements pertaining to fuel storage safety, including subcriticality requirements, and analyses demonstrate that the proposed storage patterns meet these requirements and criteria with adequate margins. Therefore, the proposed storage patterns cannot cause a new or different kind of accident.

Operation with the added weight of the Metamic inserts will not create a new or different accident. The net effect of the adding the maximum number of inserts is to add less than one percent to the weight of the loaded racks. Furthermore, the analyses of the racks with Metamic inserts installed demonstrate that the stress levels in the rack modules continue to be considerably less than allowable stress limits. Therefore, the added weight from the inserts cannot cause a new or different kind of accident.

Operation with insert movement above stored fuel will not create a new or different kind of accident. The insert with its handling tool weighs considerably less than the weight of a single fuel assembly. Single fuel assemblies are routinely moved safely over fuel assemblies and the same level of safety in design and operation will be maintained when moving the inserts. Furthermore, the effect of a dropped insert to block the top of a storage cell has been evaluated in thermal-hydraulic analyses. Therefore, the movement of inserts cannot cause a new or different kind of accident.

Whereas the installed rack inserts will displace a very small fraction of the fuel pool water volume and impose a very small reduction in operator response time to previously-evaluated SFP accidents, the reduction will not promote a new or different kind of accident. Also, displacement of water along two sides of a stored fuel assembly may have some local reduction in the peripheral cooling flow; however, this effect would be small compared to the flow induced through the fuel assembly and would in no way promote a new or different kind of accident.

The accidents and events previously analyzed and presented in the Boraflex Remedy and Alternative Source Term LARs remain bounding.

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?

No. The proposed change was evaluated for its effect on current margins of safety as they relate to criticality, structural integrity, and spent fuel heat removal capability.

The margin of safety for subcriticality required by 10 CFR 50.68(b)(4) is unchanged. New criticality analysis confirms that operation in accordance with the proposed amendment continues to meet the required subcriticality margins.

The structural evaluations for the racks and spent fuel pool with Metamic inserts installed show that the rack and spent fuel pool are unimpaired by loading combinations during seismic motion, and there is no adverse seismic-induced interaction between the rack and Metamic inserts.

The proposed change does not affect spent fuel heat generation or the spent fuel pool cooling systems. A conservative analysis indicates that the design basis requirements and criteria for spent fuel cooling continue to be met with the Metamic inserts in place, and displacing coolant. Thermal hydraulic analysis of the local effects of an installed rack insert blocking peripheral flow show a small increase in local water and fuel clad temperatures, but will remain within acceptable limits including no departure from nucleate boiling.

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

Based on the above discussion, FPL has determined that the proposed change does not involve a significant hazards consideration.

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

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

NRC Branch Chief: Douglas A. Broaddus.

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.

Virginia Electric and Power Company: Docket Nos. 50-338 and 50-339, North Anna Power Station, Unit Nos. 1 and 2, Located in Louisa County, Virginia; and 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Located in Surry County, Virginia

Date of amendment request: May 6 and February 10, 2010.

Brief description of amendment request: The proposed amendments will add Optimized ZIRLO as an acceptable fuel rod cladding material and in addition, propose adding the Westinghouse topical report for Optimized ZIRLO to the analytical methods used to determine the core operating limits listed in the Technical Specifications.

Date of publication of individual notice in Federal Register: August 27, 2010 (75 FR 52781).

Expiration date of individual notice: Comments, September 27, 2010; Hearing, October 26, 2010.

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 Start Printed Page 61529License, 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.

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: October 30, 2009, as supplemented by letters dated April 29 and August 24, 2010.

Brief description of amendment: The amendments consisted of administrative changes to update the licenses and the technical specifications as a result of changes that were approved in previously issued amendments. The amendments removed requirements that are no longer applicable due to the completion of power uprates, the replacement of steam generators, the removal of part-length control element assemblies, the completion of the core protection calculator upgrade, and made a minor administrative change to the nomenclature of the containment sump trash racks and screens.

Date of issuance: September 10, 2010.

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

Amendment No.: Unit 1-179; Unit 2-179; Unit 3-179.

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 Federal Register: January 26, 2010 (75 FR 4113). The supplemental letters dated April 29 and August 24, 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 September 10, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 19, 2009.

Brief description of amendment: The amendment revises the charcoal testing criteria in Technical Specification 5.5.9, “Ventilation Filter Testing Program.”

Date of issuance: September 13, 2010.

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

Amendment No.: 265.

Facility Operating License No. DPR-26: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: January 26, 2010 (75 FR 4115).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 13, 2010.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: September 9, 2009.

Brief description of amendment: The amendment revised Technical Specification (TS) 3/4 .9.7, “Crane Travel—Fuel Handling Building,” to permit certain operations needed for dry cask storage of spent nuclear fuel. Specifically, the proposed change to this TS, while continuing to prohibit travel of a heavy load over irradiated fuel assemblies in the spent fuel pool, would permit travel of loads in excess of 2,000 pounds over a transfer cask containing irradiated fuel assemblies, provided a single-failure-proof handling system is used.

Date of issuance: September 13, 2010.

Effective date: As of the date of issuance and shall be implemented prior to the start of the dry cask storage operations.

Amendment No.: 227.

Facility Operating License No. NPF-38: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: November 17, 2009 (74 FR 59261). The supplemental letters dated June 8 and July 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 as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 13, 2010.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October 22, 2009.

Brief description of amendment: The amendment modified the Technical Specifications (TS) Table 2.2-1 and Table 3.3-1. Specifically, the TS changes clarify TS Table 2.2-1 Notes (1) and (5), TS Table 3.3-1 Notes (a) and (c), and TS Table 3.3-1 Actions 2 and 3, which have resulted in Plant Protection System redundancy issues with respect to verbatim compliance. While the changes modified the table notations for the 10−4 percent Bistable in the Tables, they still maintain the safety function associated with the Core Protection Calculators and High Logarithmic Power trip functions, and with the small hysteresis for the 10−4 percent Bistable, there is a negligible impact on the Control Element Assembly withdrawal analysis. Additionally, the calculated peak power and heat flux are not significantly changed.

Date of issuance: September 13, 2010.Start Printed Page 61530

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

Amendment No.: 228.

Facility Operating License No. NPF-38: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: December 15, 2009 (74 FR 66384).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 13, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 29, 2010.

Brief description of amendment: The amendments revise Technical Specification (TS) 5.5.7, “Reactor Coolant Pump Flywheel Inspection Program,” to extend the reactor coolant pump (RCP) motor flywheel examination frequency from the currently-approved 10-year inspection interval to an interval not to exceed 20 years for certain Braidwood and Byron RCPs. These changes are consistent with TS Task Force (TSTF) traveler TSTF-421, “Revision to RCP Flywheel Inspection Program (WCAP-15666),” Revision 0, that has been approved generically for the Westinghouse Standard Technical Specifications, NUREG-1431. A notice announcing the availability of this proposed TS change using the Consolidated Line Item Improvement Process was published in the Federal Register on October 22, 2003 (68 FR 60422).

Date of issuance: September 16, 2010.

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

Amendment Nos.: Braidwood Unit 1-163; Braidwood Unit 2-163; Byron Unit No. 1-169; and Byron Unit No. 2-169.

Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: The amendments revise the TSs and Licenses.

Date of initial notice in Federal Register: May 18, 2010 (75 FR 27827).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 16, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 2, 2010, as supplemented by letters dated June 19, 2009, and March 31, 2010.

Brief description of amendment: The amendment revises the Exelon Nuclear Radiological Emergency Plan Annex for Clinton Station, Table B-1, “Minimum Staffing Requirements for the On-Shift Clinton Station ERO,” to increase the Non-Licensed Operator staffing from two to four, allow in-plant protective actions to be performed by personnel assigned to other functions, and replace a Mechanical Maintenance person with a Non-Licensed Operator.

Date of issuance: September 21, 2010.

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

Amendment No.: 191.

Facility Operating License No. NPF-62: The amendment revised the Facility Operating License.

Date of initial notice in Federal Register: June 1, 2010 (75 FR 30445).

The June 19, 2009, and March 31, 2010, supplement, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 27, 2010, as supplemented by letters dated May 12, and May 13, 2010.

Brief description of amendments: The amendments would revise the Operating License and technical Specifications to implement an increase of approximately 1.65 percent in rated thermal power from the current licensed thermal power of 3489 megawatts thermal (MWt) to 3546 MWt.

Date of issuance: September 16, 2010.

Effective date: As of the date of issuance and shall be implemented within 90 days for Unit 1 and within 90 days of completion of refueling outage L2R13, which is currently scheduled for March 2011, for Unit 2.

Amendment Nos.: 198/185.

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

Date of initial notice in Federal Register: May 11, 2010 (75 FR 26289).

The May 12, and May 13, 2010, supplements, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 16, 2010.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 23rd day of September 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. 2010-24815 Filed 10-4-10; 8:45 am]

BILLING CODE 7590-01-P