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 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 March 11, 2010, to March 24, 2010. The last biweekly notice was published on March 23, 2010 (75 FR 13786).

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

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

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

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

Written comments may be submitted by mail to the Chief, Rulemaking and Directives Branch (RDB), TWB-05-B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be faxed to the RDB at 301-492-3446. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.

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

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the 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 Start Printed Page 17440opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by telephone at (301) 415-1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

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

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through 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 Start Printed Page 17441or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/​EHD_​Proceeding/​home.asp, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.

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: October 29, 2009.

Description of amendment request: The amendments would delete a license condition located in each of the unit's Facility Operating Licenses (FOLs) which restricts the maximum fuel rod average burnup. Deletion of this condition would allow the maximum fuel rod average burnup to increase.

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.

Deletion of the MNS [McGuire Nuclear Station] and CNS [Catawba Nuclear Station] FOL Appendix B conditions currently limiting maximum rod average burnup to 60 GWd/MTU [Gigawatt-day per Metric Ton Uranium] does not add, delete, or modify any MNS or CNS systems, structures, or components (SSCs). The proposed amendment would effectively allow future increases in the MNS and CNS maximum rod average burnup limit up to and including 62 GWd/MTU using existing fuel management methods, analyses, and models that have been reviewed and approved by the NRC [Nuclear Regulatory Commission]. Maximum average rod burnup limits will continue to be maintained within safe and acceptable limits using these fuel management methods and models.

Increasing the MNS and CNS maximum rod average burnup limit does not affect the thermal hydraulic response or the radiological consequences of any previously evaluated accident. The fuel rod design criteria will continue to be met at the maximum burnup limits allowed utilizing the current fuel management, analysis, and evaluation processes. An increase to the maximum rod average burnup limit will not increase the likelihood of a malfunction of nuclear fuel since the fuel currently used at MNS and CNS has been designed to support a maximum rod average burnup up to and including 62 GWd/MTU. 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 amendment would delete MNS and CNS FOL Appendix B conditions which currently limits maximum rod average burnup to 60 GWd/MTU. The proposed amendment would effectively allow future increases in the MNS and CNS maximum rod average burnup limit up to and including 62 GWd/MTU using existing fuel management methods, analyses, and models that have been reviewed and approved by the NRC. The proposed amendment does not change the design function of the nuclear fuel or create any credible new failure mechanisms or malfunctions for the nuclear fuel. Fuel rod design criteria will continue to be met at the maximum burnup limits allowed under the fuel management methods and models that have been previously reviewed and approved by the NRC. 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.

The proposed amendment would delete a MNS and CNS FOL Appendix B conditions which currently limits maximum rod average burnup to 60 GWd/MTU. The proposed amendment would effectively allow future increases in the MNS and CNS maximum rod average burnup limit up to and including 62 GWd/MTU using existing fuel management methods, analyses, and models that have been reviewed and approved by the NRC. The proposed amendment does not result in altering or exceeding a design basis or safety limit for the plant. All current fuel design criteria will continue to be satisfied, and the safety analysis of record, including evaluations of the radiological consequences of design bases accidents, will remain applicable. Radiological consequences have been evaluated consistent with methodologies approved by the NRC. [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: Ms. Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, EC07H, Charlotte, NC 28202.

NRC Branch Chief: Gloria Kulesa.

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: October 29, 2009.

Description of amendment request: The amendments would delete a license condition located in each of the unit's Facility Operating Licenses (FOLs) which restricts the maximum fuel rod average burnup. Deletion of this condition would allow the maximum fuel rod average burnup to increase.

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.

Deletion of the MNS [McGuire Nuclear Station] and CNS [Catawba Nuclear Station] Start Printed Page 17442FOL Appendix B conditions currently limiting maximum rod average burnup to 60 GWd/MTU [Gigawatt-day per Metric Ton Uranium] does not add, delete, or modify any MNS or CNS systems, structures, or components (SSCs). The proposed amendment would effectively allow future increases in the MNS and CNS maximum rod average burnup limit up to and including 62 GWd/MTU using existing fuel management methods, analyses, and models that have been reviewed and approved by the NRC [Nuclear Regulatory Commission]. Maximum average rod burnup limits will continue to be maintained within safe and acceptable limits using these fuel management methods and models.

Increasing the MNS and CNS maximum rod average burnup limit does not affect the thermal hydraulic response or the radiological consequences of any previously evaluated accident. The fuel rod design criteria will continue to be met at the maximum burnup limits allowed utilizing the current fuel management, analysis, and evaluation processes. An increase to the maximum rod average burnup limit will not increase the likelihood of a malfunction of nuclear fuel since the fuel currently used at MNS and CNS has been designed to support a maximum rod average burnup up to and including 62 GWd/MTU. 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 amendment would delete MNS and CNS FOL Appendix B conditions which currently limits maximum rod average burnup to 60 GWd/MTU. The proposed amendment would effectively allow future increases in the MNS and CNS maximum rod average burnup limit up to and including 62 GWd/MTU using existing fuel management methods, analyses, and models that have been reviewed and approved by the NRC. The proposed amendment does not change the design function of the nuclear fuel or create any credible new failure mechanisms or malfunctions for the nuclear fuel. Fuel rod design criteria will continue to be met at the maximum burnup limits allowed under the fuel management methods and models that have been previously reviewed and approved by the NRC. 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.

The proposed amendment would delete a MNS and CNS FOL Appendix B conditions which currently limits maximum rod average burnup to 60 GWd/MTU. The proposed amendment would effectively allow future increases in the MNS and CNS maximum rod average burnup limit up to and including 62 GWd/MTU using existing fuel management methods, analyses, and models that have been reviewed and approved by the NRC. The proposed amendment does not result in altering or exceeding a design basis or safety limit for the plant. All current fuel design criteria will continue to be satisfied, and the safety analysis of record, including evaluations of the radiological consequences of design bases accidents, will remain applicable. Radiological consequences have been evaluated consistent with methodologies approved by the NRC. [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: Ms. Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, EC07H, Charlotte, NC 28202.

NRC Branch Chief: Gloria Kulesa.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of amendment request: February 8, 2010.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) requirements related to TS 3.1.3, “Control Rod Operability,” and TS 3.1.5, “Control Rod Scram Accumulators,” to be consistent with NUREG-1433, “Standard Technical Specifications General Electric Plants, BWR/4.” The proposed amendment also corrects certain typographical errors.

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 changes involve an administrative change to LCO [limiting condition for operation] 3.1.3, “Control Rod OPERABILITY,” and a simplification in the modeling methodology for scram time analysis in LCO 3.1.5, “Control Rod Scram Accumulators,” that continue to ensure that control rod operability requirements for the number and distribution of operable, slow and stuck control rods satisfy scram reactivity rate assumptions used in the plant safety analysis.

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 changes do not involve any physical alteration of the plant (no new or different type of equipment is being installed) and do not involve a change in the design, normal configuration, or basic operation of the plant. The proposed changes do not introduce any new accident initiators. The proposed changes do not involve significant changes in the fundamental methods governing normal plant operation and do not require unusual or uncommon operator actions. The proposed changes provide assurance that the plant will not be operated in a mode or condition that violates the assumptions or initial conditions in the safety analyses and that the systems, structures, and components (SSCs) remain capable of performing their intended safety functions as assumed in the same analyses. Consequently, the response of the plant and the plant operator to postulated events will not be significantly different.

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

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

Response: No.

Margin of safety is related to confidence in the ability of fission product barriers to perform their intended design functions during and following an accident. The proposed changes address control rod operability and continue to ensure control rod scram time acceptance criteria is satisfied. The scram time test acceptance criteria and control rod operability restrictions are based on industry approved methodology and will continue to ensure control rod scram design functions and reactivity insertion assumptions used in the safety analyses continue to be protected.

Therefore, the proposed changes do 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: William A. Horin, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Michael T. Markley.

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: January 28, 2010.

Description of amendment request: The proposed license amendment Start Printed Page 17443request modifies the licensee's commitment to Table B-1, “Minimum Staffing Requirements for NRC Licensees for Nuclear Power Plant Emergencies,” of NUREG-0654/FEMA-REP-1, Revision 1, “Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,” dated November 1980. Current Table 13.3-17, “Repair and Corrective Actions,” of the Emergency Plan only allows that Electrical or Instrumentation & Control technicians may fill these two positions. This change will allow these two maintenance positions on shift to be filled with any combination of the three maintenance craft disciplines.

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 not involve a significant increase in the probability or consequences of an accident previously evaluated.

No.

The proposed change does not increase the probability or consequences of an accident. The change only impacts the implementation of the Emergency Plan by changing staffing of the Repair and Corrective action functions after an event. It has no impact on plant equipment or the operation of plant equipment and thus has no impact on the probability or consequences of an event. The number of personnel on shift has not been revised from the current Emergency Plan. The repair and corrective action function would continue to be performed by trained personnel because the process, personnel, and equipment involved in implementing the Emergency Plan would complete the same functions as those completed under the existing Emergency Plan, the Plan would continue to ensure adequate protection of public health and safety.

(2) Does not create the possibility of a new or different kind of accident from any accident previously evaluated.

No.

The change only impacts the implementation of the Emergency Plan by changing staffing of the Repair and Corrective action functions after an event. The change does not impact any plant equipment or systems needed to respond to an accident, nor does it involve any analysis of plant accidents. The proposed change does not create a new or different kind of accident from any previously evaluated because this change only impacts emergency response repair functions.

(3) Does not involve a significant reduction in a margin of safety.

No.

The change to the Emergency Plan does not reduce the margin of safety currently provided by the Plan as it maintains the current number of personnel on shift to perform Repair and Corrective action functions. Repair and corrective actions will continue to be performed by trained personnel. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: January 24, 2010.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) Section 1.0, Definitions, TS Section 3.6, Primary System Boundary Specifications 3.6.A, and TS Administrative Controls Section 5.5, to include reference to the Pressure and Temperature Limits Report (PTLR). The PTLR includes revised 34 effective full-power years (EFPY) P-T Curves, neutron fluence, and Adjusted Reference Temperature (ART) values.

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 modifies Technical Specifications (TS) Section 1.0

(“Definitions”), Specification 3.6.A.2, and revises 5.0 (“Administrative Controls”), to include section 5.5.9 to include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of SIR-05-044-A, “Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,” dated April-2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF [Technical Specification Task Force] -419-A (“Revised PTLR Definition and References in ISTS 5.6.6, RCS [reactor coolant system] PTLR”). In an NRC Safety Evaluation [safety evaluation] Report dated February 6, 2007, “the NRC staff has found that SIR-05-044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors to the extent,” specified and under, the limitations delineated in the TR and in the enclosed final SE.” As part of this change, the Pilgrim Pressure and Temperature Limits Report (PTLR) based on the methodology and template provided in SIR-05-044-A is being supplied for review. The pressure and temperature curves utilize the methodology of SIR-05-044-A.

The NRC has established requirements in Appendix G to 10 CFR [Part] 50 in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. Additionally, the regulation in 10 CFR Part 50, Appendix H, provides the NRC staff's criteria for the design and implementation of RPV material surveillance programs for operating light water reactors. Implementing this NRC approved methodology does not reduce the ability to protect the reactor coolant pressure boundary as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components. Incorporation of the new methodology for calculating P-T curves, and the relocation of the P-T curves from the TS to the PTLR provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions. 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 does not affect the assumed accident performance of the RCPB, nor any plant structure, system, or component previously evaluated. The proposed change does not involve the installation of new equipment, and installed equipment is not being operated in a new or different manner. The change in methodology ensures that the RCPB remains capable of performing its safety functions. No set points are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced which could introduce the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change does not affect the function of the RCPB or its response during plant transients. There are no changes proposed which alter the set points at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation. This change adopts the methodology of SIR-05-044-A, “Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,” dated April 2007 for preparation of the pressure and temperature curves. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

This change adopts the methodology of SIR-05-044-A, “Pressure-Temperature Start Printed Page 17444Limits Report Methodology for Boiling Water Reactors,” dated April 2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A (“Revise PTLR Definition and References in [Improved Standard Technical Specification] ISTS 5.6.6, RCS PTLR”). In an NRC Safety Evaluation Report dated February 6, 2007, the NRC staff has found that SIR-05-044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors.”

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.

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: December 3, 2009.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) to incorporate Standard Technical Specification 3.1.8 “Scram Discharge Volume (SDV) Vent and Drain Valves” and 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. The operation of Vermont Yankee Nuclear Power Station (VY) in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed amendment does not impact the operability of any structure, system or component that affects the probability of an accident or that supports mitigation of an accident previously evaluated. The proposed amendment does not affect reactor operations or accident analysis and has no radiological consequences. The operability requirements for accident mitigation systems remain consistent with the licensing and design basis. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The operation of VY in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

3. The operation of VY in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.

The proposed change ensures that the safety functions of the SDV vent and drain valves are fulfilled. The isolation function is maintained by valves in the vent and drain lines and by the required action to isolate the affected line. The ability to vent and drain the SDVs is maintained through administrative controls. In addition, the reactor protection system ensures that an SDV will not be filled to the point that it has insufficient volume to accept a full scram. Maintaining the safety functions related to isolation of the SDV and insertion of control rods ensures that the proposed change does not involve a significant reduction in the margin of safety. The proposed amendment does not change the design or function of any component or system. The proposed amendment does not impact any safety limits, safety settings or safety margins. Therefore, operation of VY in accordance with the proposed amendment will not involve a significant reduction in the margin to 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.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine Mile Point Nuclear Station Unit No. 2 (NMP 2), Oswego County, New York

Date of amendment request: December 9, 2009.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.8.4, “DC Sources—Operating,” by removing the Mode restrictions for performance of TS Surveillance Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division 3 direct current (DC) electrical power subsystem battery. These surveillances verify that the battery capacity is adequate for the battery to perform its required functions. The proposed amendment would remove these Mode restrictions for the Division 3 battery, thereby allowing performance of SR 3.8.4.7 and SR 3.8.4.8 for the Division 3 battery during Mode 1, 2, or 3 in conjunction with scheduled high pressure core spray (HPCS) system outages. Eliminating the requirement to perform SR 3.8.4.7 and SR 3.8.4.8 during Mode 4 or 5 (cold shutdown or refueling conditions) will provide greater flexibility in scheduling Division 3 battery testing activities by allowing the testing to be performed during non-outage times.

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 Division 3 (HPCS) DC electrical power subsystem and its associated emergency loads are accident mitigating features, not accident initiators. Therefore, the proposed TS changes to allow performance of Division 3 battery surveillance testing (service test and the battery performance discharge test) in any plant operating mode will not significantly impact the probability of any previously evaluated accident.

The design and function of plant equipment is not being modified by the proposed amendment. Neither the battery test frequency nor the time that the TSs allow the HPCS system to be inoperable are being revised. Battery testing in accordance with the proposed TS changes will continue to verify that the Division 3 DC electrical power subsystem is capable of performing its required function of providing DC power to HPCS system equipment, consistent with the plant safety analyses. The battery testing period is within the period of time that the HPCS system will already be out of service for a planned system outage. The battery testing does not increase unavailability of the supported HPCS system or represent any change in risk above the current practice of planned system maintenance outages. Any risk associated with the testing of the Division 3 battery will be enveloped by the risk management of the HPCS system outage. In addition, the HPCS system reliability and availability are monitored and evaluated in relationship to Maintenance Rule goals to ensure that total outage times do not degrade operational safety over time.

Testing is limited to only one electrical division of equipment at a time to ensure that design basis requirements are met. Should a fault occur while testing the Division 3 battery, there would be no significant impact on any accident consequences since the other two divisional DC electrical power subsystems and their associated emergency loads would be available to provide the minimum safety functions necessary to shut Start Printed Page 17445down 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 battery surveillance testing is conducted. Performance of these surveillance tests while online will continue to verify operability of the Division 3 battery. The proposed license amendment does not impact any plant systems that are accident initiators and does not adversely impact any accident mitigating systems, since the HPCS system will already be out of service. The battery testing will not increase the out-of-service time for the HPCS system.

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 battery do not affect the operability requirements for the battery, as verification of such operability will continue to be performed as required. Continued verification of operability supports the capability of the Division 3 DC electrical power subsystem to perform its required function of providing DC 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.

The battery testing will be performed when the HPCS system is already out of service for a planned system outage. The battery testing does not increase unavailability of the supported HPCS system or represent any change in risk above the current practice of planned system maintenance outages, as currently allowed by the TS.

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: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006.

NRC Branch Chief: Nancy L. Salgado.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine Mile Point Nuclear Station Unit No. 2 (NMP 2), Oswego County, New York

Date of amendment request: December 18, 2009.

Description of amendment request: The proposed amendment would modify Technical Specifications (TS) requirements for unavailable barriers by adding limiting condition for operation (LCO) 3.0.9. The NRC staff issued a Notice of Opportunity to Comment in the Federal Register on June 2, 2006 (71 FR 32145), on possible amendments to revise the plant-specific TSs, including a model safety evaluation and model no significant hazards consideration 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 October 3, 2006 (71 FR 58444). The licensee affirmed the applicability of the model no significant hazards consideration determination in its application dated December 18, 2009.

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 allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed LCO 3.0.9 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.9. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to an unavailable barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG [Regulatory Guide] 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.9 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP [Incremental Conditional Core Damage Probability] and ICLERP [Incremental Conditional Large Early Release Probability]) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation published in the Federal Register on October 3, 2006. Therefore, this change does not involve a Start Printed Page 17446significant reduction in a margin of safety.

The NRC staff has reviewed the 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: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006.

NRC Branch Chief: Nancy L. Salgado.

Northern States Power Company—Minnesota, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: October 27, 2009.

Description of amendment request: The proposed amendment would adopt the Alternative Source Term (AST) methodology, in addition to Technical Specification (TS) changes supported by the AST design basis accident radiological consequences analyses. The proposed amendment would also incorporate Technical Specification Task Force (TSTF)-490, “Deletion of E-Bar Definition and Revision to RCS [reactor coolant system] Specific Activity Tech Spec,” Revision 0.

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

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

With this change, Prairie Island Nuclear Generating Plant (PINGP) proposes to implement 10 CFR 50.67, alternative source term methodologies, implement approved industry improved Standard Technical Specification traveler, TSTF-490, and revise TS 3.3.7, “Spent Fuel Pool Special Ventilation System Actuation Instrumentation,” TS 3.7.12, “Auxiliary Building Special Ventilation System,”

TS 3.7.13, “Spent Fuel Pool Special Ventilation System,” TS 3.9.4, “Containment Penetrations,” TS 5.5.9, “Ventilation Filter Testing Program,” TS 5.5.14, “Containment Leakage Rate Testing Program,” and TS 5.5.16, “Control Room Habitability Program.”

Alternative source term (AST) calculations have been performed for PINGP that demonstrate the dose consequences are consistent with the regulatory limits of 10 CFR 50.67 and the guidance of Regulatory Guide (RG) 1.183. The use of the AST methodology changes the regulatory assumptions regarding the analytical treatment of the design basis accidents and has no direct effect on the probability of any accident. AST methods have been utilized in the analysis of the limiting design basis accidents, as follows: loss of coolant accident, fuel handling accident, main steam line break, steam generator tube rupture, control rod ejection accident, and locked rotor accident. The results of the analyses, which include the proposed changes to the Technical Specifications, demonstrate that the dose consequences of these limiting events are within regulatory limits.

Reactor coolant specific activity is not an initiator for any accident previously evaluated. The Completion Time when reactor coolant gross activity is not within limit is not an initiator for any accident previously evaluated. The current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. As a result, the proposed change does not significantly increase the probability of an accident. The proposed change will limit reactor coolant noble gases to concentrations consistent with the accident analyses. The proposed change to the Completion Time has no impact on the consequences of any design basis accident since the consequences of an accident during the extended Completion Time are the same as the consequences of an accident during the current Completion Time. As a result, the consequences of any accident previously evaluated are not significantly increased.

The Spent Fuel Pool Special Ventilation System is no longer credited for filtration or isolation. The Containment Penetrations TS is being replaced with a TS on Decay Time, which requires that recently irradiated fuel (<50 hours) cannot be handled. The Ventilation Filter Testing Program TS is being revised to reflect changes to filter testing. As a result of these TS changes, the probability or consequences of an accident previously evaluated are not significantly increased.

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

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

With this change, PINGP proposes to implement 10 CFR 50.67, alternative source term methodologies, implement approved industry improved Standard Technical Specification traveler, TSTF-490, and revise TS 3.3.7, “Spent Fuel Pool Special Ventilation System Actuation Instrumentation,” TS 3.7.12, “Auxiliary Building Special Ventilation System,” TS 3.7.13, “Spent Fuel Pool Special Ventilation System,” TS 3.9.4, “Containment Penetrations,” TS 5.5.9, “Ventilation Filter Testing Program,” TS 5.5.14, “Containment Leakage Rate Testing Program,” and TS 5.5.16, “Control Room Habitability Program.”

The AST methodology is not an accident initiator, as it is a method used to estimate resulting accident doses. The proposed operation of plant systems affected by this change does not create the possibility of a new or different kind of accident previously evaluated. Changes that are proposed to plant equipment (ventilation systems) pertain to accident mitigation equipment. The operation or mis-operation of these ventilation systems do not initiate any accidents. The radiological consequence analyses demonstrate that the proposed changes are acceptable. The results of the analyses, which include the proposed changes to the Technical Specifications, demonstrate that the dose consequences of these limiting events are within regulatory limits.

The proposed change in specific activity limits does not alter any physical part of the plant nor does it affect any plant operating parameter. The change does not create the potential of a new or different kind of accident from any accident previously evaluated.

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

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

With this change, PINGP proposes to implement 10 CFR 50.67, alternative source term methodologies, implement approved industry improved Standard Technical Specification traveler, TSTF-490, and revise TS 3.3.7, “Spent Fuel Pool Special Ventilation System Actuation Instrumentation,” TS 3.7.12, “Auxiliary Building Special Ventilation System,” TS 3.7.13, “Spent Fuel Pool Special Ventilation System,” TS 3.9.4, “Containment Penetrations,” TS 5.5.9, “Ventilation Filter Testing Program,” TS 5.5.14, “Containment Leakage Rate Testing Program,” and TS 5.5.16, “Control Room Habitability Program.”

The proposed implementation of the AST methodology is consistent with RG 1.183. The radiological consequences of these accidents are within the regulatory acceptance criteria associated with the use of the AST methodology. The doses at the exclusion area and low population zone boundaries and in the control room are consistent with the regulatory limits of 10 CFR 50.67 and the guidance of RG 1.183. The margin of safety for the radiological consequences of these accidents is considered to be that provided by meeting the applicable regulatory limits, which are set at or below 10 CFR 50.67 limits.

The proposed change to revise the limits on noble gas radioactivity in the primary coolant is consistent with the assumptions in the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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: Peter M. Glass, Assistant General Counsel, Xcel Energy Start Printed Page 17447Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: January 29, 2010.

Description of amendment request: The amendments would change an Emergency Action Level (EAL) scheme based on NUREG-0654, “Criteria for Preparation and Evaluation of Radiological Emergency Response Plan and Preparedness in Support of Nuclear Power Plants,” to one based on NEI 99-01, “Methodology for Development of Emergency Action Levels,” Revision 4. This would change the methodology for deriving selected Notification of Unusual Event values in Table R-1, Gaseous Effluent Monitor Classification Thresholds, and deleting EAL RA2.4 which evaluates abnormal radiation readings at infrequently accessed areas and revise the radiation level threshold values for Reactor Coolant System (RCS) letdown indication.

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.

These changes affect the North Anna [* * *] Power Station Emergency Action Levels, but do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes have no effect on the consequences of any analyzed accident since the changes do not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or consequence 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.

These changes affect the North Anna [* * *] Power Station Emergency Action Levels, but do not alter any of the requirements of the Operating License or the Technical Specifications. They do not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No system setpoints are being modified. No new failure modes are introduced by the proposed changes. The proposed amendment does not introduce accident initiator or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3:

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

Response: No.

These changes affect the North Anna [* * *] Power Station Emergency Action Levels, but do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not affect any of the assumptions used in the accident analysis, nor do they affect any operability requirements for equipment important to plant safety. Therefore, the proposed changes will not result in a significant reduction in the margin of safety as defined in the bases for technical specifications covered in this license amendment request. [Therefore, this 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Gloria Kulesa.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: January 27, 2010.

Description of amendment request: The proposed license amendment request would increase each unit's rated power (RP) level from 2546 megawatts thermal (MWt) to 2587 MWt, and make Technical Specifications changes as necessary to support operation at the uprated power level. The proposed change is an increase in RP of approximately 1.6%.

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 consequence of an accident previously evaluated?

Response: No.

The proposed change will increase the Surry Power Station (SPS) Units 1 and 2 rated power (RP) from 2546 megawatts thermal (MWt) to 2587 MWt. Nuclear steam supply system and balance-of-plant systems, components and analyses that could be affected by the proposed change to the RP were evaluated using revised design parameters. The evaluations determined that these structures, systems and components are capable of performing their design function at the proposed uprated RP of 2587 MWt. An evaluation of the accident analyses demonstrates that the applicable analysis acceptance criteria are still met with the proposed changes. Power level is an input assumption to equipment design and accident analyses, but it is not a transient or accident initiator. Accident initiators are not affected by the power uprate, and plant safety barrier challenges are not created by the proposed changes.

The radiological consequences of operation at the uprated power conditions have been assessed. The proposed change to RP does not affect release paths, frequency of release, or the analyzed reactor core fission product inventory for any accidents previously evaluated in the SPS Updated Final Safety Analysis Report. There is a small increase in the reactor coolant activity concentration. Structures, systems and components required to mitigate transients are capable of performing their design functions with the proposed changes, and are thus acceptable. Analyses performed to assess the effects of mass and energy releases remain valid. The assessment of radiological consequences for operation at the proposed power level confirmed that there is not a significant increase for affected events.

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.

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of any proposed changes. The ultrasonic flow meter (UFM) being installed to facilitate the Measurement Uncertainty Recapture (MUR) power uprate has been analyzed, and system failures will not adversely affect any safety-related system or any structures, systems or components required for transient mitigation. Structures, systems and components previously required for transient mitigation are still capable of fulfilling their intended design functions. The proposed changes have no significant adverse affect on any safety-related structures, systems or components and do not significantly change the performance or integrity of any safety-related system.

The proposed changes do not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than previously evaluated. Operating at an RP of 2587 MWt does not create any new accident initiators or precursors. Credible Start Printed Page 17448malfunctions are bounded by the current accident analyses of record or recent evaluations demonstrating that applicable criteria are still met with the proposed changes.

Therefore, this 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 margins of safety associated with the power uprate are those pertaining to core thermal power. These include fuel cladding, reactor coolant system pressure boundary, and containment barriers. Core analyses demonstrate that power uprate implementation will continue to meet the current nuclear design basis. Impacts to components associated with the reactor coolant system pressure boundary structural integrity, and factors such as pressure-temperature limits, vessel fluence, and pressurized thermal shock were determined to be bounded by the current analyses.

Systems will continue to operate within their design parameters and remain capable of performing their intended safety functions following implementation of the proposed change. The current SPS safety analyses, and the revised design basis radiological accident dose calculations, bound the power uprate without significantly impacting margins.

Therefore, this 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: Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.

NRC Branch Chief: Gloria Kulesa.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: January 29, 2010.

Description of amendment request: The amendments would change an Emergency Action Level (EAL) scheme based on NUREG-0654, “Criteria for Preparation and Evaluation of Radiological Emergency Response Plan and Preparedness in Support of Nuclear Power Plants,” to one based on NEI 99-01, “Methodology for Development of Emergency Action Levels,” Revision 4. This would change the methodology for deriving selected Notification of Unusual Event values in Table R-1, Gaseous Effluent Monitor Classification Thresholds, and deleting EAL RA2.4 which evaluates abnormal radiation readings at infrequently accessed areas.

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.

These changes affect the [* * *] Surry Power Station Emergency Action Levels, but do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes have no effect on the consequences of any analyzed accident since the changes do not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or consequence 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.

These changes affect the [* * *] Surry Power Station Emergency Action Levels, but do not alter any of the requirements of the Operating License or the Technical Specifications. They do not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No system setpoints are being modified. No new failure modes are introduced by the proposed changes. The proposed amendment does not introduce accident initiator or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3:

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

Response: No.

These changes affect [* * *] the Surry Power Station Emergency Action Levels, but do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not affect any of the assumptions used in the accident analysis, nor do they affect any operability requirements for equipment important to plant safety. Therefore, the proposed changes will not result in a significant reduction in the margin of safety as defined in the bases for technical specifications covered in this license amendment request. [Therefore, this 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Gloria Kulesa.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: December 16, 2009.

Description of amendment request: The proposed changes would revise Technical Specification (TS) 3.8.4, “DC [Direct Current] Sources—Operating,” Surveillance Requirement (SR) 3.8.4.2 and SR 3.8.4.5 to revise the battery connection resistance acceptance criteria for inter-cell connections from ≤ 150E-6 ohms to ≤ 33E-6 ohms and would add connection resistance acceptance criteria for inter-tier connections and inter-bank connection of ≤ 150E-6 ohms.

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 to revise the SR 3.8.4.2 and SR 3.8.4.5 acceptance criteria for battery connection resistance will not challenge the ability of the safety-related batteries to perform their safety function. Appropriate monitoring and maintenance will continue to be performed on the safety related batteries. Current TS testing and monitoring requirements will not be altered.

The proposed change does not involve a physical change to the batteries, nor does it change the safety function of the batteries. The proposed TS revision involves no significant changes to the operation of any systems or components in normal and accident operating conditions and no changes to existing structures, systems or components.

Therefore, this change will not increase the probability or consequences of an accident previously evaluated.

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

Response: No

The proposed changes to revise the SR 3.8.4.2 and SR 3.8.4.5 acceptance criteria for battery connection resistance is an increase in conservatism, without a change in system Start Printed Page 17449testing methods, operation, or control. Safety related batteries installed in the plant will be required to meet criteria more restrictive and conservative than current acceptance criteria and standards. The proposed change does not affect the manner in which the batteries are tested and maintained, thus there are no new failure mechanisms for the system.

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

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

Response: No

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes will not adversely affect operation of plant equipment, as the changes being made are more restrictive. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new battery maintenance and monitoring program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety related loads in accordance with analysis assumptions.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, N.W., Washington, DC 20037.

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.

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: October 2, 2008.

Brief description of amendment request: The proposed amendment would revise the Technical Specifications (TS) associated with the verification of ice condenser door operability and TS surveillance requirements 3.6.13.5 and 3.6.13.6.

Date of publication of individual notice in Federal Register: March 8, 2010 (75 FR 10513).

Expiration date of individual notice: Comments April 7, 2010; Hearing May 7, 2010.

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: October 2, 2008.

Brief description of amendment request: The proposed amendment would revise the Technical Specifications (TS) associated with the verification of ice condenser door operability and TS surveillance requirements 3.6.13.5 and 3.6.13.6.

Date of publication of individual notice in Federal Register: March 8, 2010 (75 FR 10508).

Expiration date of individual notice: Comments April 7, 2010; Hearing May 7, 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 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 email to pdr.resource@nrc.gov.

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: June 19, 2009, as supplemented by letter dated October 20, 2009.

Brief description of amendment: The proposed amendment would revise Technical Specification 3.3.1, “Reactor Protection System Instrumentation.” The proposed change revises the requirements related to the reactor protection system interlock for the turbine trip input to the reactor protection system.

Date of issuance: March 17, 2010.

Effective date: Effective as of the date of issuance and shall be implemented by the end of Refueling Outage 26.

Amendment No.: 222.

Renewed Facility Operating License No. DPR-23: The amendment revises the technical specifications.

Date of initial notice in Federal Register: January 5, 2010 (75 FR 460).

The Commission's related evaluation of the amendment is contained in a safety evaluation dated March 17, 2010.Start Printed Page 17450

Public comments received as to proposed no significant hazards consideration (NSHC): No.

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

Date of amendment request: March 13, 2008, as supplemented by letter dated February 28, 2010.

Brief description of amendment: The amendment replaced the current ANO-1 Technical Specification 3.4.12, “RCS [Reactor Coolant System] Specific Activity,” limit on RCS gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent Xe-133 definition that would replace the current E Bar average disintegration energy definition. In addition, the current dose equivalent I-131 definition would be revised to allow the use of additional thyroid dose conversion factors.

Date of issuance: March 18, 2010.

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

Amendment No.: 243.

Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications/license.

Date of initial notice in Federal Register: May 6, 2008 (73 FR 25038). The supplemental letter dated February 28, 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 March 18, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 2, 2009, as supplemented by letter dated June 24, 2009.

Brief description of amendment: The amendment modified Technical Specification (TS) 3.3.1.1, “Reactor Protective Instrumentation,” and TS 3.3.2.1, “Engineered Safety Feature Actuation System Instrumentation,” specifically, Table 3.3-1, Table 4.3-1, and Table 3.3-3, to adopt a mode of applicability for the Logarithmic Power Level—High, Pressurizer Pressure—Low, Steam Generator [SG] Pressure—Low, and the SG Differential Pressure and Level Low functions. These changes are consistent with NUREG-1432, Revision 3.0, “Standard Technical Specifications, Combustion Engineering Plants,” dated June 2004.

Date of issuance: March 11, 2010.

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

Amendment No.: 289.

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

Date of initial notice in Federal Register: June 2, 2009 (74 FR 26433). The supplemental letter dated June 24, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on June 2, 2009 (74 FR 26433).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 11, 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 19, 2009.

Brief description of amendment: The amendment relocated the Waterford 3 Steam Generator Level—High trip requirements from Technical Specification Sections 2.2 and 3/4.3.1 to the Technical Requirements Manual (TRM). This change is consistent with Technical Specification Task Force (TSTF) 410-A, “Relocation of Steam Generator Level—High Trip to the TRM,” and Revision 3 of NUREG-1432, “Standard Technical Specifications, Combustion Engineering Plants.”

Date of issuance: March 18, 2010.

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

Amendment No.: 225.

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

Date of initial notice in Federal Register: December 1, 2009 (74 FR 62834).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 18, 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: December 4, 2008, as supplemented by letters dated February 17, 2009; July 27, 2009; December 4, 2009; and January 29, 2010.

Brief description of amendment: The amendments revise Technical Specifications (TSs) 1.1, “Definitions,” and 3.4.16, “RCS [Reactor Coolant System] Specific Activity,” and Surveillance Requirements 3.4.16.1, 3.4.16.2, and 3.4.16.3. The revisions replace the current TS 3.4.16 limit on RCS gross specific activity with a new limit on RCS noble gas-specific activity. The revisions adopt TS Task Force (TSTF) Change Traveler, TSTF-490, “Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec [sic],” Revision 0.

Date of issuance: March 23, 2010.

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

Amendment Nos.: Braidwood Unit 1-162; Braidwood Unit 2-162; Byron Unit No. 1-167; and Byron Unit No. 2-167.

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: January 27, 2009 (74 FR 4771).

The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 23, 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: March 26, 2009, as supplemented by letter dated October 28, 2009.

Brief description of amendments: The proposed changes would revise Technical Specification 3.5.1, “Emergency Core Cooling Systems (ECCS) Operating,” to delete the existing allowance with the Automatic Depressurization System accumulator backup compressed gas system that Start Printed Page 17451currently allows a completion time of 72 hours to restore bottle pressure to ≥ 500 psig.

Date of issuance: March 19, 2010.

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

Amendment Nos.: 196/183.

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

Date of initial notice in Federal Register: September 8, 2009 (74 FR 46242). The October 28, 2009 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 amendments is contained in a Safety Evaluation dated March 19, 2010.

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: November 6, 2008, supplemented by letters dated December 11, 2008, July 2, 2009, October 2, 2009, and November 24, 2009.

Brief description of amendment: The amendment replaces the current TMI-1 technical specification limit on Reactor Coolant System (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit is based on a new dose equivalent Xenon-133 definition that replaces the previous E-Bar average disintegration energy definition. In addition, the dose equivalent Iodine-131 definition has been revised.

Date of issuance: March 11, 2010.

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

Amendment No.: 272.

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

Date of initial notice in Federal Register: March 10, 2009 (74 FR 10309). The supplements dated December 11, 2008, July 2, 2009, October 2, 2009, and November 24, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 11, 2010.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 11, 2009, as supplemented by letters dated August 12 and December 21, 2009, and March 5, 2010.

Brief description of amendment: The amendment revised Surveillance Requirements 3.8.4.2 and 3.8.4.5 in Technical Specification Section 3.8.4, “DC [Direct Current] Sources—Operating,” by adding a parameter of total battery resistance to the values of battery connection resistance.

Date of issuance: March 18, 2010.

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

Amendment No.: 236.

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

Date of initial notice in Federal Register: May 5, 2009 (74 FR 20752). The supplemental letters dated August 12 and December 21, 2009, and March 5, 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 March 18, 2010.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit No. 2 (NMP2), Oswego County, New York

Date of application for amendment: March 30, 2009, as supplemented on November 2, 2009.

Brief description of amendment: The amendment modifies the NMP2 Technical Specification (TS) 3.8.1, “AC Sources—Operating,” to remove operating mode restrictions for the performance of certain Surveillance Requirements (SRs) pertaining to the Division 3, High Pressure Core Spray (HPCS) Emergency Diesel Generator (DG). The testing in Modes 1 or 2 were previously prohibited in SR 3.8.1.7, SR 3.8.1.8, and SR 3.8.1.10, and in Modes 1, 2, or 3 in SR 3.8.1.9, SR 3.8.1.11, SR 3.8.1.14, SR 3.8.1.15, and SR 3.8.1.17. The amendment removes these Mode restrictions and allows the above SRs to be performed in any operating mode for the Division 3 DG. The Mode restrictions remain applicable to the other two safety-related (Division 1 and Division 2) DGs.

Date of issuance: March 18, 2010.

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

Amendment No.: 133.

Renewed Facility Operating License No. NPF-069: The amendment revises the License and TSs.

Date of initial notice in Federal Register: June 16, 2009 (74 FR 28577).

The supplemental letter dated November 2, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 18, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 27, 2009.

Description of amendment request: The amendments revised the Technical Specifications to change Surveillance Requirement 3.6.1.3, “Primary Containment Isolation Valves,” to eliminate unnecessary local leak rate tests.

Date of issuance: March 22, 2010.

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

Amendment Nos.: 277, 304, and 263.

Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: Amendments revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: October 20, 2009 (74 FR 53781).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 22, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 20, 2009, as supplemented by letters dated December 10, 2009, and January 19, 2010.

Brief description of amendment: The amendment revised Technical Specification (TS) 5.5.16, “Containment Leakage Rate Testing Program.” The revision reflects a one-time extension of Start Printed Page 17452the current containment Type A leak rate test (integrated leak rate test or ILRT) interval requirement of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” Option B, “Performance Based Requirements,” from 10 years to 15 years. The amendment allows the next ILRT to be performed no later than October 25, 2014.

Date of issuance: March 17, 2010.

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

Amendment No.: 195.

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

Date of initial notice in Federal Register: August 25, 2009 (74 FR 42931). The supplemental letters dated December 10, 2009, and January 19, 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 March 17, 2010.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 25th day of March 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. 2010-7451 Filed 4-5-10; 8:45 am]

BILLING CODE 7590-01-P