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

As Enacted ( Federal Register)

Cited authorities 9

Cited in

Federal Register: March 9, 2010 (Volume 75, Number 45)

Notices

Page 10823-10833

From the Federal Register Online via GPO Access [wais.access.gpo.gov]

DOCID:fr09mr10-83

NUCLEAR REGULATORY COMMISSION

NRC-2010-0081

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

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 February 11, 2010, to February 24, 2010. The last biweekly notice was published on February 23, 2010 (75 FR 8139).

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

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subject facility operating license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the

Commission's ``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at

One White Flint North, Public File Area O1F21, 11555 Rockville Pike

(first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's

(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the

Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic

Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/ petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

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

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the submitter an e-mail notice confirming receipt of the document. The

E-Filing system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the agency's adjudicatory E-

Filing system may seek assistance by contacting the NRC Meta System

Help Desk through the ``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at

MSHD.Resource@nrc.gov, or by a toll-free call at (866) 672-7640. The

NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,

Eastern Time, Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:

Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth

Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in

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

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

CFR 2.309(c)(1)(i)-(viii).

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint

North, Public File Area O1F21, 11555 Rockville Pike (first floor),

Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC PDR Reference staff at 1-800- 397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,

Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: May 28, 2009.

Description of amendment request: The amendments would revise

Technical Specification (TS) 3.8.1, ``AC Sources-Operating,'' to restrict voltage limits for the applicable TS 3.8.1 surveillances governing the Emergency Diesel Generators (EDGs).

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The increase in the minimum EDG output voltage acceptance value in TS 3.8.1 Surveillance Requirements does not adversely affect any of the parameters in the accident analyses. The proposed change increases the minimum allowed EDG output voltage to ensure that sufficient voltage is available to operate the required

Emergency Safety Feature (ESF) equipment under accident conditions.

Additionally the increase in minimum voltage output voltage allowed ensures that adequate voltage is available to support the assumptions made in the Design Bases Accident (DBA) analyses. This conservative change of the EDG voltage output acceptance criteria does not affect the probability of evaluated accidents, but rather provides increased assurance that the EDGs will provide a sufficient voltage. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The increase in the minimum EDG output voltage acceptance criterion supports the assumptions in the accident analyses that sufficient voltage will be available to operate ESF equipment on the

Class 1E buses when these buses are powered from the Emergency

Diesel Generators. The maximum EDG output voltage of 4580 volts is not affected by this change. The change in minimum output voltage from 3740 to 3950 volts ensures the reliability of the onsite emergency power source. Therefore, the proposed change will 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 margin of safety?

This proposed license amendment is limited to increasing the minimum EDG output voltage acceptance criterion in TS 3.8.1

Surveillance Requirements. No other surveillance criterion is affected. The surveillance frequencies and test requirement are unchanged. This amendment provides increased assurance that the EDG will provide sufficient voltage to its respective components to ensure design requirements are satisfied. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this 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.

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NRC Branch Chief: Gloria Kulesa.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,

Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: July 1, 2009.

Description of amendment request: The proposed amendments would revise TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation'' and TS 1.1, ``Definitions.'' The proposed amendments support plant modifications which would replace the existing Source Range (SR) and

Intermediate Range (IR) excore detector systems with equivalent neutron monitoring systems. The new instrumentation will perform both the SR and the IR monitoring functions.

Implementation of the above changes will entail plant modifications and will impact the Updated Final Safety Analysis Reports (UFSAR). The necessary UFSAR revisions will be submitted in accordance with 10 CFR 50.71(e).

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 Technical Specification changes are in support of a plant modification involving the replacement and upgrade of the

Nuclear Instrumentation System (NIS) Source Range and Intermediate

Range instrumentation. The specific Technical Specification changes are associated with 1) the methods of calibrating NIS channels; 2) the definition of Nominal Trip Setpoint; 3) the specific Nominal

Trip Setpoint and Allowable Values for various NIS channels, including the Intermediate Range, Source Range and Intermediate

Range Permissive ``P-6'' instrumentation; 4) the addition of specific requirements to be taken if an as-found Intermediate Range or Source Range channel setpoint is outside its predefined as-found tolerance; and 5) the addition of specific requirements regarding resetting of an Intermediate Range or Source Range channel setpoint within an as-left tolerance.

The NIS is accident mitigation equipment and does not affect the probability of any accident being initiated. In addition, none of the above-mentioned proposed Technical Specification changes affect the probability of any accident being initiated.

The performance of the replacement SR and IR detectors and associated equipment will equal or exceed that of the existing instrumentation. The proposed changes to Nominal Trip Setpoints and

Allowable Values are based on accepted industry standards and will preserve assumptions in the applicable accident analyses. None of the proposed changes alter any assumption previously made in the radiological consequences evaluations, nor do they affect mitigation of the radiological consequences of an accident previously evaluated.

In summary, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of any of the proposed changes.

The NIS is not capable by itself of initiating any accident. Other than the replacement of the detectors themselves and the associated hardware, no physical changes to the overall plant are being proposed. No changes to the overall manner in which the plant is operated are being proposed. Therefore, none of the proposed changes will create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions.

These barriers include the fuel cladding, the reactor coolant system pressure boundary, and the containment barriers. The modification to replace the SR and IR detectors and associated equipment will not have any impact on these barriers. In addition, the proposed

Technical Specification changes will not have any impact on these barriers. No accident mitigating equipment will be adversely impacted as a result of the modification. The proposed changes do not affect any safety analysis conclusions because the SR and IR neutron flux trips are not explicitly credited in any accident analysis. The replacement instrumentation will have overall performance capabilities equal to or greater than those for the existing instrumentation. Therefore, existing safety margins will be preserved. None of the proposed changes will involve a significant reduction in a margin of safety.

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: July 1, 2009.

Description of amendment request: The proposed amendments would revise TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation.'' The proposed amendments support plant modifications which would replace the existing Source Range (SR) and Intermediate Range (IR) excore detector systems with equivalent neutron monitoring systems. The new instrumentation will perform both the SR and the IR monitoring functions.

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 Technical Specification changes are in support of a plant modification involving the replacement and upgrade of the

Nuclear Instrumentation System (NIS) Source Range and Intermediate

Range instrumentation. The specific Technical Specification changes are associated with (1) the methods of calibrating NIS channels; (2) the definition of Nominal Trip Setpoint; (3) the specific Nominal

Trip Setpoint and Allowable Values for various NIS channels, including the Intermediate Range, Source Range and Intermediate

Range Permissive ``P-6'' instrumentation; (4) the addition of specific requirements to be taken if an as-found Intermediate Range or Source Range channel setpoint is outside its predefined as-found tolerance; and (5) the addition of specific requirements regarding resetting of an Intermediate Range or Source Range channel setpoint within an as-left tolerance.

The performance of the replacement SR and IR detectors and associated equipment will equal or exceed that of the existing instrumentation. The proposed changes to Nominal Trip Setpoints and

Allowable Values are based on accepted industry standards and will preserve assumptions in

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the applicable accident analyses. None of the proposed changes alter any assumption previously made in the radiological consequences evaluations, nor do they affect mitigation of the radiological consequences of an accident previously evaluated.

Response: No

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of any of the proposed changes.

Response: No

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions.

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. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

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

Carolina

Date of amendment request: May 18, 2009.

Description of amendment request: The proposed amendments would revise the Technical Specifications to adopt Technical Specification

Task Force (TSTF) Standard Technical Specification Change Traveler

TSTF-248. TSTF 248 modifies the definition of shutdown margin.

Criterion 1:

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

Response: No.

The revision to SDM [shutdown margin] definition will result in analytical flexibility for determining SDM. Changes in the definition will not have an impact on the probability of an accident previously evaluated.

The introduction of this definition change does not change continued compliance with all applicable regulatory requirements and design criteria (e.g., train separation, redundancy, and single failure). Therefore, since all plant systems will continue to function as designed, all plant parameters will remain within their design limits. As a result, the proposed changes will not increase the consequences of an accident.

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

Criterion 2:

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

Response: No.

Revising the TS [Technical Specifications] definition of SDM would not require core designers to revise any SDM boron calculations. Rather, it would afford the analytical flexibility for determining SDM for a particular circumstance.

The proposed changes do not involve any change in the design, configuration, or operation of the nuclear plant. The current plant safety analyses, therefore, remain complete and accurate in addressing the design basis events and in analyzing plant response and consequences.

The Limiting Conditions for Operations, Limiting Safety System

Settings and Safety Limits specified in the Technical Specifications are not affected by the proposed changes. As such, the plant conditions for which the design basis accident analyses were performed remain valid.

The amendment does not introduce a new mode of plant operation or new accident precursors, does not involve any physical alterations to plant configurations or make changes to system set points that could initiate 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.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their accident mitigation functions. These barriers include the fuel and fuel cladding, the reactor coolant system, and the containment and containment related systems. The proposed changes will not impact the reliability of these barriers to function. Radiological doses to plant operators or to the public will not be impacted as a result of the proposed change. The change in the TS definition will have no impact to these barriers. Adequate SDM will continue to be ensured for all operational conditions.

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

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. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

Carolina

Date of amendment request: August 6, 2009.

Description of amendment request: The proposed amendments would revise the Technical Specifications by changing the surveillance requirement for the low temperature overpressure protection system

(LTOP) from 6 months to 18 months.

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

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issue of no significant hazards consideration, which is presented below:

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

No. This is a revision to the Technical Specification (TS)

Surveillance Requirement (SR) for performing the channel calibration for the power operated relief valve (PORV). As such, the TS SR interval extension continues to ensure the calibration is performed in a time frame supported by current analysis. The instrumentation loop has been upgraded to an environmentally qualified instrumentation loop with improved instrument uncertainty and reliability. The accidents previously evaluated have not changed.

Therefore, extending the TS SR frequency from 6 months to 18 months does not significantly increase the probability or consequences of any accident previously evaluated.

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

No. This revision does not impact the LTOP evaluation analysis.

The method for testing remains the same. The proposed SR frequency is supported by an environmentally qualified instrumentation loop with improved instrument uncertainty and reliability.

Therefore, extending the TS SR frequency from 6 months to 18 months will not create the possibility of a new or different kind of accident from any kind of accident previously evaluated.

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

No. The proposed change does not adversely affect any plant safety limits, setpoints, or design parameters. The change also does not adversely affect the fuel, fuel cladding, Reactor Coolant

System, or Containment Operability.

Therefore, extending the TS SR frequency from 6 months to 18 months does not involve a significant reduction in a margin of safety.

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. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

Carolina

Date of amendment request: August 31, 2009.

Description of amendment request: The proposed amendments would revise the Technical Specifications to allow one of the two required 230kV switchyard 125 VDC power source batteries to be inoperable for up to 10 hours for the purpose of replacing an entire battery bank and performing the required testing.

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

No. This License Amendment Request (LAR) proposes to permit one of the two 230 kV switchyard 125 VDC batteries to be out of service for up to ten days when it is necessary to replace and test a complete battery (all cells of one battery bank). The capacity of each battery, needing only 58 of 60 cells to be available (i.e., two cells can be jumpered out), is sufficient to carry the loads of both distribution centers during replacement.

The 230kV switchyard 125 VDC power system is credited to provide uninterruptible power to specified loads during certain design basis events. The probability of any of these events occurring is not impacted by removing one of the batteries for replacement. The consequences associated with permitting a 230 kV switchyard 125 VDC battery to be out of service for up to ten days for battery replacement have been evaluated. The likelihood of an event occurring during the additional time a battery bank will be out of service is essentially the same as that of an event occurring during the 24 hour period permitted by the existing completion time.

Operation in accordance with the amendment authorizing this change would not involve any accident initiation sequences or radiological release pathways that could affect the consequences of any accident analyzed. Use of this additional time for battery replacement will be infrequent since battery replacement normally is performed at or near the end of the twenty year qualified life.

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

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

No. This License Amendment Request (LAR) proposes to permit one of the two 230 kV switchyard 125 VDC batteries to be out of service for up to ten days when it is necessary to replace and test a complete battery (all cells of one battery). Operation in accordance with this proposed amendment will not result in any new plant equipment, alter the present plant configuration, nor adversely affect how the plant is currently operated.

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

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

Since the proposed change will not physically alter the present plant configuration nor adversely affect how the plant is currently operated, the proposed change does not adversely affect any plant safety limits, setpoints, or design parameters. The change also does not adversely affect the fuel, fuel cladding, Reactor Coolant System or containment integrity.

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

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. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

Carolina

Date of amendment request: September 30, 2009.

Description of amendment request: The proposed amendments would revise the Technical Specifications to allow performance of testing containment spray nozzles for nozzle blockage following activities which could result in nozzle blockage, rather than a fixed periodic basis. Currently the testing for nozzle blockage is performed every 10 years.

Criterion 1:

Does the proposed change involve a significant increase in the probability or

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consequences of an accident previously evaluated?

No. The proposed amendment will modify CNS [Catawba Nuclear

Station] SR [surveillance requirement] 3.6.6.7, MNS [McGuire Nuclear

Station] SR 3.6.6.7, and ONS [Oconee Nuclear Station] SR 3.6.5.8 to change the frequency for verifying spray nozzles are unobstructed.

The proposed change modifies the frequency for performance of a surveillance test which does not impact any failure modes that could lead to an accident. The proposed frequency change does not affect the ability of the spray nozzles or spray system to perform its accident mitigation function as assumed and therefore there is no effect on the consequence of any accident. Verification of no blockage continues to be required, but now verification will be performed following activities that could result in nozzle blockage.

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?

No. The proposed amendment will modify CNS SR 3.6.6.7, MNS SR 3.6.6.7, and ONS SR 3.6.5.8 to change the frequency for verifying spray nozzles are unobstructed. The spray systems are not being physically modified and there is no impact on the capability of the system to perform accident mitigation functions. No system setpoints are being modified and no changes are being made to the method in which borated water is delivered to the spray nozzles. The testing requirements imposed by this proposed change to check for nozzle blockage following activities that could cause nozzle blockage do not introduce new failure modes for the system. The proposed amendment does not introduce accident initiators 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?

No. The proposed amendment will modify CNS SR 3.6.6.7, MNS SR 3.6.6.7, and ONS SR 3.6.5.8 to change the frequency for verifying spray nozzles are unobstructed. The proposed change does not change or introduce any new setpoints at which mitigating functions are initiated. No changes to the design parameters of the spray systems are being proposed. There are no changes in system operation being proposed by this change that would impact an established safety margin. The proposed change modifies the frequency for verification of nozzle operability in such a way that continued high confidence exists that the spray systems will continue to function as designed.

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

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.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: November 19, 2009, as supplemented by letter dated January 28, 2010.

Description of amendment request: The proposed change will modify the test acceptance criteria in Surveillance Requirement (SR) 3.8.1.10 for the Diesel Generator endurance surveillance test. The proposed change will also incorporate changes to the Standard Technical

Specifications made by Technical Specification Task Force (TSTF) 238-A,

Revision 3 and TSTF-276-A, Revision 2. Specifically, the proposed change will modify SR notes in TS 3.8.1 and TS 3.8.4

No. The proposed changes revise the acceptance criteria to be applied to an existing surveillance test of the facility emergency diesel generators (EDGs), allows deviation from that acceptance criteria for certain grid conditions, and allows testing in modes that is normally not done. Performing a surveillance test is done under conditions where it is not an accident initiator and does not increase the probability of an accident occurring. The proposed new acceptance criteria will assure that the EDGs are capable of carrying the peak electrical loading assumed in the various existing safety analyses which take credit for the operation of the EDGs.

Establishing acceptance criteria that bound existing analyses validates the related assumption used in those analyses regarding the capability of equipment to mitigate accident conditions. The deviation allowed for grid conditions does not affect the capability of the testing to achieve these purposes. The proposed change to allow testing in modes normally restricted requires an evaluation to ensure, prior to performing the test, that the potential consequences are capable of being addressed by existing procedures and does not create transients or conditions that could significantly affect the possibility of an accident. Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The proposed changes revise the test acceptance criteria for a specific performance test conducted on the existing EDG, allows deviation from that acceptance criteria for certain grid conditions, and allows testing in modes that is normally not done. The proposed changes do not involve installation of new equipment or modification of existing equipment, so no new equipment failure modes are introduced. The proposed revision to the EDG surveillance test acceptance criteria also is not a change to the way that the equipment or facility is operated and no new accident initiators are created. The proposed testing on line must be evaluated to assure plant safety is maintained or enhanced, inherent in such an evaluation would be that the testing does not create the possibility of a new or different kind of accident. Therefore the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety?

No. The conduct of performance tests on safety-related plant equipment is a means of assuring that the equipment is capable of maintaining the margin of safety established in the safety analyses for the facility. The proposed change in the EDG technical specification surveillance test acceptance criteria is consistent with values assumed in existing safety analyses and is consistent with the design rating of the EDGs. The allowance for certain grid conditions does not alter this conclusion since the power factors are conservatively determined. Testing allowed in modes when it is not normally performed is limited to conditions where an evaluation is performed to assure plant safety is maintained or enhanced.

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

Attorney for licensee: Mr. William C. Dennis, Assistant General

Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White

Plains, NY 10601.

NRC Branch Chief: Nancy L. Salgado.

Page 10830

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,

Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: December 18, 2009.

Description of amendment request: The proposed amendment would incorporate the use of alternate methodologies for the calculation of reactor pressure vessel beltline weld initial reference temperatures, the calculation of the adjusted reference temperatures (ARTs), the development of the reactor pressure vessel pressure-temperature (P-T) limit curves, and the low temperature reactor coolant system (RCS) overpressure analysis into Technical Specification (TS) 5.6.4. The amendment would also revise the analysis requirement for the low temperature RCS overpressure events from 21 to 32 Effective Full Power

Years (EFPY) contained in Operating License (OL) Condition 2.C(3)(d).

An application that addressed similar issues was previously submitted on April 15, 2009, and the notice of that application was provided in the Federal Register on June 16, 2009 (72 FR 28577). Since the licensee eliminated one of the alternate methodologies for the calculation of the adjusted reference temperature (as described in the April 15, 2009, application) and replacing it with the existing Nuclear Regulatory

Commission (NRC)-approved methodology, which is described in Regulatory

Guide 1.99, Revision 2, ``Radiation Embrittlement of Reactor Vessel

Materials'', in December 19, 2009, the application is being renoticed in its entirety. The notice supersedes the notice published in the

Federal Register on June 16, 2009.

Response: No.

The amendment request proposes two changes to the TS/OL. The first change incorporates the use of alternative methodologies to develop the [Davis-Besse Nuclear Power Station, Unit No. 1] DBNPS P-

T limit curves and [low temperature over pressure] LTOP limits into

TS 5.6.4 to augment the existing listed methodology of BAW-10046A,

Revision 2. The second change revises OL Condition 2.C(3)(d) to reflect the revised LTOP analysis is valid to 32 [Effective Full

Power Years] EFPY.

The first change incorporates the use of Topical Report BAW- 2308, Revisions 1-A and 2-A and [American Society of Mechanical

Engineers] ASME Code Cases N-588 and N-640. The topical report and

ASME code cases have been approved or accepted for use by the NRC

(provided that any conditions/limitations are satisfied). The proposed additions to the methodologies for the reactor vessel P-T curve and LTOP limit development provide an acceptable means of satisfying the requirements of 10 CFR 50, Appendix G. The proposed additions do not alter the design, function, or any operation of any plant equipment. Therefore, the proposed additions do not affect the probability or consequences of any previously evaluated accidents, including reactor coolant pressure boundary failures.

The second change is considered administrative in nature and reflects the revised methodologies. It will not alter the design, function, or operation of any plant equipment. Therefore, the proposed change does not affect the probability or consequences of any previously evaluated accidents. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The amendment request proposes two changes to the TS/OL. The first change incorporates the use of alternative methodologies to develop the DBNPS P-T limit curves and LTOP limits into TS 5.6.4 to augment the existing listed methodology of BAW-10046A, Revision 2.

The second change revises OL Condition 2.C(3)(d) to reflect that the revised analysis is valid to 32 EFPY.

The first change incorporates methodologies that either have been approved or accepted for use by the NRC (provided that any conditions/limitations are satisfied). The changes do not alter the design, function, or operation of any plant equipment. The P-T limit curves and LTOP limits will provide the same level of protection to the reactor coolant boundary as was previously evaluated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The second change is considered administrative in nature and reflects the revised methodologies. It will not alter the design or operation of any plant equipment. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The second change revises OL Condition 2.C(3)(d) to reflect that the revised analysis is valid to 32 EFPY. The first change incorporates methodologies that either have been approved or accepted for use by the NRC (provided that any conditions/limitations are satisfied).

The second change is considered administrative in nature and reflects the revised methodologies. The changes do not alter the design, function, or operation of any plant equipment. The P-T limit curves and LTOP limits will provide the same level of protection to the reactor coolant boundary as was previously evaluated. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy

Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Stephen Campbell.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of amendment request: November 30, 2009.

Description of amendment request: The proposed amendment would modify conditions and associated actions to Technical Specification 3.8.1, ``AC [Alternating Current] Sources Operating.'' The proposed amendment would revise the Completion Time for restoring one or more inoperable diesel generators (DGs) in one train to an operable status and increase the Completion Time for confirming that the other DGs are not impacted by a common cause failure. 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 diesel generators (DGs) are designed as backup alternating current (ac) power sources in the event of loss of offsite power.

The proposed changes to Completion Times associated with determining inoperable DGs are not subject to common cause failure and restoration of inoperable DGs and the deletion of the note referencing the C-S DG do not change the conditions, operating configurations, or minimum amount of operating equipment assumed in the safety analysis accident mitigation. No changes are proposed in the manner in which the DGs provide plant protection.

Page 10831

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes associated with determining inoperable DGs are not subject to common cause failure and restoration of inoperable DGs and the deletion of the note referencing the C-S DG do not involve a change in design, configuration, or method of operation of the plant. The proposed changes will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. The capability of the DGs to perform their required safety function will not be affected. The proposed changes do not affect the interaction of the

DGs with any system whose failure or malfunction can initiate an accident. As such, no new failure modes are being introduced.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The DGs are designed as backup AC power sources in the event of loss of offsite power. The proposed changes associated with determining inoperable DGs are not subject to common cause failure and restoration of inoperable DGs and the deletion of the note referencing the C-S DG do not change conditions, operating configurations, or minimum amount of operating equipment assumed in the safety analysis accident mitigation. The proposed changes do not alter the plant design, including instrument setpoints, nor do they alter the assumptions contained in the safety analyses. No changes are proposed in the manner in which the DGs provide plant protection or which create new modes of plant operation.

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: L. Raghavan.

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 change would revise the approved fire protection program as described in the Wolf Creek

Generating Station (WCGS) Updated Safety Analysis Report (USAR) to allow use of the fire-resistive cable for certain power and control cables associated with two motor-operated valves on Train B Component

Cooling Water System. This will be a deviation from certain technical commitments to Title 10 of the Code of Federal Regulations (10 CFR)

Part 50, Appendix R, Section III.G.2, as described in Appendix 9.5E of the WCGS USAR.

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 design function of structures, systems and components are not impacted by the proposed change. The proposed change involves the use of fire-resistive cable at WCGS for certain power and control cables associated with two motor-operated valves (EGHV0016 and EGHV0054) on Train B Component Cooling Water System and will not initiate an event. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The

Meggitt Si 2400 fire-resistive cable has been independently tested to applicable requirements and the implementation design reflects the test results. Therefore, the probability of any accident previously evaluated is not increased. Equipment required to mitigate an accident remains capable of performing the assumed function.

Response: No.

The proposed change will not alter the requirements or function for systems required during accident conditions. The design function of structures, systems and components are not impacted by the proposed change. No new or different accidents result from implementing Meggitt Si 2400 fire-resistive cable in Fire Areas A-16 and A-21. The Meggitt Si 2400 fire-resistive cable has been independently tested to applicable requirements and the implementation design reflects the test results. The use of Meggitt

Si 2400 fire-resistive cable is not a significant change in the methods governing normal plant operation.

Response: No.

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

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

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw

Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

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 change would revise the approved fire protection program as described in the Wolf Creek

Cooling Water System. This will be a deviation from certain technical commitments to Title 10 of the Code of Federal Regulations (10 CFR)

Part 50, Appendix R, Section III.G.2, as described in Appendix 9.5E of the WCGS USAR.

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?

Page 10832

Response: No.

Si 2400 fire-resistive cable is not a significant change in the methods governing normal plant operation.

Response: No.

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

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

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw

Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint

North, Public File Area 01F21, 11555 Rockville Pike (first floor),

Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System (ADAMS) Public

Electronic Reading Room on the internet at the NRC Web site, http:// www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr.resource@nrc.gov.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point

Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: March 29, 2009, as supplemented by letters dated September 21 and December 22, 2009.

Brief description of amendment: The amendment established a more restrictive acceptance criterion for surveillance requirement (SR) 3.8.6.6 regarding periodic verification of capacity for the affected station batteries.

Date of issuance: February 24, 2010.

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

Amendment No.: 264.

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

License and the Technical Specifications.

Date of initial notice in Federal Register: May 19, 2009 (74 FR 23444). The supplemental letters dated September 21 and December 22, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 24, 2010.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendment: February 25, 2009.

Brief description of amendment: The changes remove the provisions contained in Technical Specification (TS) 3/4.4.8, which specify requirements relating to the structural integrity of American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components. This specification is redundant to the requirements contained within Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, ``Codes and standards.'' With this change, the pressure boundary structural integrity of ASME Code Class 1, 2 and 3

Page 10833

components will continue to be maintained through the facility's compliance with 10 CFR 50.55a.

Date of issuance: February 24, 2010.

Effective date: As of the date of issuance and shall be implemented within 180 days of issuance.

Amendment Nos.: 199 and 160.

Facility Operating License Nos. NPF-39 and NPF-85. These amendments revised the license and the technical specifications.

Date of initial notice in Federal Register: April 21, 2009 (74 FR 18254).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 24, 2010.

No significant hazards consideration comments received: No.

National Aeronautics and Space Administration, Docket Nos. 50-30, and 50-185. Erie County, Ohio

Date of amendment request: January 9, 2009, as supplemented by letter dated October 6, 2009.

Brief description of amendment: The amendment adds a condition to each license requiring that the National Aeronautics and Space

Administration assess the residual radioactivity and demonstrate that the stream bed and banks of Plum Brook between the Plum Brook Station boundary and Sandusky Bay meet the radiological criteria for unrestricted use specified in 10 CFR 20.1402 prior to terminating

Licenses TR-3 and R-93.

Date of issuance: February 1, 2010.

Effective date: February 1, 2010.

Amendment Nos.: 14 and 10, respectively.

Possession Only License Nos. TR-3 and R-93: The amendment revises both licenses.

Date of initial notice in Federal Register: May 5, 2009 (74 FR 20751).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation Report, dated February 1, 2010.

No Significant Hazards Consideration Comments Received: No.

National Aeronautics and Space Administration, Docket Nos. 50-30, and 50-185. Erie County, Ohio (TAC NO. J00301)

Date of amendment request: January 9, 2009, as supplemented by letter dated October 6, 2009.

Brief description of amendment: The amendment adds a condition to each license requiring that the National Aeronautics and Space

Licenses TR-3 and R-93.

Date of issuance: February 1, 2010.

Effective date: February 1, 2010.

Amendment Nos.: 14 and 10, respectively

Possession Only License Nos. TR-3 and R-93: The amendment revises both licenses.

Date of initial notice in Federal Register: May 5, 2009 (74 FR 20751)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation Report, dated February 1, 2010.

No Significant Hazards Consideration Comments Received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear

Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of application for amendments: April 9, 2009.

Brief description of amendments: The amendments relocate Technical

Specification (TS) requirements pertaining to communications during refueling operations (TS 3/4.9.5), manipulator crane operability (TS 3/ 4.9.6), and crane travel (TS 3/4.9.7) to the Technical Requirements

Manual.

Date of issuance: February 17, 2010.

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

Amendment Nos.: 293 and 277.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the TSs and the License.

Date of initial notice in Federal Register: August 25, 2009 (74 FR 42929).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 17, 2010.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 25th day of February 2010.

For the Nuclear Regulatory Commission.

Allen G. Howe,

Deputy Director, Division of Operating Reactor Licensing, Office of

Nuclear Reactor Regulation.

FR Doc. 2010-4523 Filed 3-8-10; 8:45 am

BILLING CODE 7590-01-P

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

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Thousands of Data Sources

Find What You Need, Quickly

Over 2 million registered users

You can sign up for a trial and make the most of our service including these benefits.

Why Sign-up to vLex?

Over 100 Countries

Thousands of Data Sources

Find What You Need, Quickly

Over 2 million registered users

You can sign up for a trial and make the most of our service including these benefits.

Why Sign-up to vLex?

Over 100 Countries

Thousands of Data Sources

Find What You Need, Quickly

Over 2 million registered users

You can sign up for a trial and make the most of our service including these benefits.

Why Sign-up to vLex?

Over 100 Countries

Thousands of Data Sources

Find What You Need, Quickly

Over 2 million registered users

Why Sign-up to vLex?

Over 100 Countries

Thousands of Data Sources

Find What You Need, Quickly

Over 2 million registered users

You can sign up for a trial and make the most of our service including these benefits.

Why Sign-up to vLex?

Over 100 Countries

Thousands of Data Sources

Find What You Need, Quickly

Over 2 million registered users