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Notice

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

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Information about this document as published in the Federal Register.

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I. Background

Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from March 24, 2011, to April 6, 2011. The last biweekly notice was published on April 5, 2011 (76 FR 18801).

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. Start Printed Page 21918Under the Commission's regulations in Title 10 of the Code of Federal Regulations (10 CFR), § 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

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

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

Written comments may be submitted by mail to the Chief, Rules, Announcements and Directives Branch (RADB), TWB-05-B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be faxed to the RADB at 301-492-3446. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 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, then 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 Start Printed Page 21919(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 identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

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

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

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the agency's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html, by e-mail at MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, 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://ehd1.nrc.gov/​EHD/​, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 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.

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit 1 and 2, Calvert County, Maryland

Date of amendments request: October 25, 2010.

Description of amendments request: The amendment would revise Technical Specification Limiting Condition for Operation (LCO) 3.0.5 to provide clarification as to when the LCO can be invoked in order to perform required testing to demonstrate OPERABILITY of equipment.Start Printed Page 21920

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 change to LCO 3.0.5 more clearly specifies the situations when LCO 3.0.5 can be applied. In some Technical Specifications, the steps taken to comply with ACTIONS involve the placement of redundant or alternate equipment or trains into service, or the repositioning (e.g., opening or closing) or components. The proposed change would allow the use of LCO 3.0.5 in situations such as these. This proposed change does not, however, change the intent of LCO 3.0.5. The purpose of LCO 3.0.5 remains to provide an exception to LCO 3.0.2, to not comply with the applicable Required Action(s) while performing required testing to demonstrate the OPERABILITY of either equipment being returned to service or the OPERABILITY of other equipment.

The proposed change does not affect any analyzed accident initiators, nor does it change the units' ability to successfully respond to any previously evaluated accident. As a result, there is also no change to existing radiological assumptions used in the accident evaluations. In addition this proposed change does not change the operation or maintenance performed on operating equipment.

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 change to LCO 3.0.5 more clearly specifies the situations when LCO 3.0.5 can be applied. In some Technical Specifications, the steps taken to comply with ACTIONS involve the placement of redundant or alternate equipment or trains into service, or the repositioning (e.g., opening or closing) or components. The proposed change would allow the use of LCO 3.0.5 in situations such as these. This proposed change does not, however, change the intent of LCO 3.0.5. The purpose of LCO 3.0.5 remains to provide an exception to LCO 3.0.2, to not comply with the applicable Required Action(s) while performing required testing to demonstrate the OPERABILITY of either equipment being returned to service or the OPERABILITY of other equipment.

The proposed change does not involve a modification to the physical configuration of the units nor does it involve any change in the methods governing normal plant operation. The proposed change does not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally there is no change in the types or increase in the amounts of any effluent that may be released offsite and there is no increase in individual or cumulative occupational exposure.

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

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

Response: No.

The proposed change to LCO 3.0.5 more clearly specifies the situations when LCO 3.0.5 can be applied. In some Technical Specifications, the steps taken to comply with ACTIONS involve the placement of redundant or alternate equipment or trains into service, or the repositioning (e.g., opening or closing) or components. The proposed change would allow the use of LCO 3.0.5 in situations such as these. This proposed change does not, however, change the intent of LCO 3.0.5. The purpose of LCO 3.0.5 remains to provide an exception to LCO 3.0.2, to not comply with the applicable Required Action(s) while performing required testing to demonstrate the OPERABILITY of either equipment being returned to service or the OPERABILITY of other equipment.

The proposed change does not involve any modification to the physical configuration of the operating units and does not alter equipment operation. As such, the safety functions of plant equipment and their response to any analyzed accident scenario are unaffected by this proposed change and thus there is no reduction in the margin of safety.

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

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 amendments request involves no significant hazards consideration.

Attorney for licensee: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.

NRC Branch Chief: Nancy L. Salgado.

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit 1 and 2, Calvert County, Maryland

Date of amendments request: March 22, 2011.

Description of amendments request: The proposed amendment would revise the Technical Specifications (TSs) to define a new time limit for restoring inoperable reactor coolant system (RCS) leakage detection instrumentation to operable status. The proposed TS changes are consistent with TS Task Force (TSTF)-513, “Revise PWR [pressurized-water reactor] Operability Requirements and Actions for RCS Leakage Instrumentation.”

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

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

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS required operable RCS leakage detection instrumentation monitor is the containment atmosphere gaseous radiation monitor. The monitoring of RCS leakage is not a precursor to any accident previously evaluated. The monitoring of RCS leakage is not used to mitigate the consequences of any accident previously evaluated. The plant specific variation to this license amendment request, to insert the Note “Not required until 12 hours after establishment of steady state operation” into applicable portions of the Technical Specification is administrative in nature. As a result, its inclusion does not impact any plant equipment's ability to perform its required functions. Therefore, it is concluded that the proposed changes do 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 clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS required operable RCS leakage detection instrumentation monitor is the containment atmosphere gaseous radiation monitor. 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 normal plant operation. The proposed change maintains sufficient continuity and diversity of leak detection capability that the probability of piping evaluated and approved for leak-before-break progressing to pipe rupture remains extremely low. The plant specific variation to this license amendment request, to insert the Note “Not required until 12 hours after establishment of steady state operation” into applicable portions of the Technical Specification also does not involve a physical alteration of the plant or change in how plant equipment is operated. Therefore, it is concluded that the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.Start Printed Page 21921

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

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS required operable RCS leakage detection instrumentation monitor is the containment atmosphere gaseous radiation monitor. Reducing the amount of time the plant is allowed to operate with only the containment atmosphere gaseous radiation monitor operable increases the margin of safety by increasing the likelihood that an increase in RCS leakage will be detected before it potentially results in gross failure. The plant specific variation to this license amendment request, to insert the Note “Not required until 12 hours after establishment of steady state operation” into applicable portions of the Technical Specification provides clarification as it reflects the time necessary for plant conditions to stabilize in order to ensure an accurate water inventory can be obtained.

Therefore, it is concluded that 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 amendments request involves no significant hazards consideration.

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 amendments request involves no significant hazards consideration.

Attorney for licensee: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: October 20, 2010.

Description of amendment request: The proposed amendment would revise the technical specifications (TS) description of fuel assemblies specified in TS 4.2.1. Additionally, changes are requested to the analytical methods referenced in TS 5.6.5.b. The changes to TS 5.6.5.b includes the addition of AREVA topical reports, BAW-10240(P)(A), “Incorporation of M5TM Properties in Framatome ANP Approved Methods,” and EMF-2328(P)(A), “PWR Small Break LOCA Evaluation Model S-RELAP5 Based,” and the deletion of nine analytical methods that were previously approved but are no longer planned to be used, and therefore have not been analyzed for acceptability for M5TM (M5) alloy fuel.

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

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

The proposed license amendment adds a Nuclear Regulatory Commission approved analytical method, BAW-10240(P)(A), “Incorporation of M5TM Properties in Framatome ANP Approved Methods,” used to determine the core operating limits, to Technical Specification (TS) 5.6.5.b and changes the description of fuel assemblies specified in TS 4.2.1 to allow use of the M5 alloy. The proposed amendment does not affect the acceptance criteria for any Final Safety Analysis Report (FSAR) safety analysis analyzed accidents or anticipated operational occurrences. The proposed amendment does not involve operation of the required structures, systems or components (SSCs) in a manner different from those previously recognized or evaluated. As such, the proposed amendment does not increase the probability or consequences of an accident.

In addition, the proposed license amendment adds NRC approved methodology EMF-2328(P)(A), “PWR Small Break LOCA Evaluation Model, S-RELAP5 Based.” This change, by itself, does not impact the current design bases. The proposed change enables the use of new methodologies to re-analyze small break loss-of-coolant accidents. Revised analyses may either result in continued conformance within design bases, or may change the design bases. If design bases changes result from a revised analysis, then the specific design changes will be evaluated in accordance with HBRSEP, Unit 2, design change procedures and 10 CFR 50.59. Further, this part of the change does not involve physical changes to any plant structure, system, or component.

In addition, the proposed license amendment deletes nine analytical methods that were previously approved and listed in Section 5.6.5.b, but are no longer planned to be used. This change is administrative in nature as it removes methodologies that have become obsolete and hence have not been analyzed for acceptability with M5 fuel.

Therefore, operation of the facility in accordance with the proposed amendment would 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 Previously Evaluated.

Use of M5 fuel will not result in changes in the operation or configuration of the facility. Topical reports BAW-10227(P)(A) and BAW-10240(P)(A) evaluate the material properties of the M5 alloy and conclude that they are similar or better than those of zircaloy-4. Therefore, M5 fuel rod cladding will perform similarly to those fabricated from zircaloy-4, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident. No new failure mechanisms will be introduced by the changes being requested.

The proposed addition of EMF-2328(P)(A) does not involve any physical alteration of plant systems, structures, or components, other than allowing for fuel design in accordance with NRC-approved methodologies. No new or different equipment is being installed. No installed equipment is being operated in a different manner. There is no change to the parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions. As a result, no new failure modes are being introduced by introduction of this methodology.

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

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

The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the M5 alloy are not significantly different from those of zircaloy-4. M5 alloy is expected to perform similarly or better than zircaloy-4 for all normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. The proposed changes do not affect the acceptance criteria for any FSAR safety analysis analyzed accidents or anticipated operational occurrences. All required safety limits would continue to be analyzed using methodologies approved by the Nuclear Regulatory Commission.

There is no impact on any margin of safety resulting from the incorporation of these new topical reports into the Technical Specifications. If design basis changes result from a revised analysis that uses these new methodologies, the specific design changes will be evaluated in accordance with HBRSEP, Unit 2, design change procedures and 10 CFR 50.59. Any potential reduction in the margin of safety would be evaluated for that specific design change.

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

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

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Branch Chief: Douglas A. Broaddus.Start Printed Page 21922

Carolina Power and Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of amendment request: January 13, 2011.

Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) to change the description of fuel assemblies specified in TS 5.3.1 and add the AREVA NP Inc., topical report, BAW-10240(P)(A), “Incorporation of M5TM Properties in Framatome ANP Approved Methods,” to the referenced analytical methods in administrative TS 6.9.1.6.2 to allow the use of M5TM alloy for fuel rod cladding.

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 license amendment adds a NRC approved analytical method, BAW-10240(P)(A), “Incorporation of M5TM Properties in Framatome ANP Approved Methods,” used to determine the core operating limits, to TS 6.9.1.6.2 and changes the description of fuel assemblies specified in TS 5.3.1 to allow use of the M5TM alloy. The proposed amendment does not affect the acceptance criteria for any Final Safety Analysis Report (FSAR) safety analysis analyzed accidents and anticipated operational occurrences. As such, the proposed amendment does not increase the probability or consequences of an accident. The proposed amendment does not involve operation of the required structures, systems or components (SSCs) in a manner or configuration different from those previously recognized or evaluated. Therefore, operation of the facility in accordance with the proposed amendment would 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.

Use of M5TM clad fuel will not result in changes in the operation or configuration of the facility. Topical Report BAW-10240 describes, by reference, that the material properties of the M5TM alloy are similar to or better than those of Zircaloy-4. Therefore, since M5TM fuel rod cladding will perform similarly to those fabricated from Zircaloy-4, the possibility of the fuel becoming an accident initiator and causing a new or different type of accident is precluded. Since the material properties of M5TM alloy are similar to or better than those of Zircaloy-4, there will be no significant changes in the types of any effluents that may be released off-site. There will not be a significant increase in occupational or public radiation exposure. The proposed amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the changes being requested. 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 change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the M5TM alloy are not significantly different from those of Zircaloy-4. M5TM alloy is expected to perform similarly to or better than Zircaloy-4 for all normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. The proposed changes do not affect the acceptance criteria for any FSAR safety analysis analyzed accidents or anticipated operational occurrences. All required safety limits will continue to be analyzed using methodologies approved by the NRC. Therefore, the proposed amendment would not involve a significant reduction in a margin of safety. margin of safety.

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

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: January 20, 2011.

Description of amendment request: The proposed change would revise H. B. Robinson Steam Electric Plant Technical Specifications (TS) Section 3.8.3, “Diesel Fuel Oil and Starting Air,” and Section 3.8.5, “DC Sources—Shutdown.” The proposed change to TS 3.8.3 revises a nonconservative air receiver tank pressure to a value consistent with vendor recommendations. The proposed change to TS 3.8.5 corrects an editorial error related to TS formatting.

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

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

The proposed change to TS 3.8.3 revises a non-conservative value in the current TS for EDG air start pressure. The proposed value is consistent with vendor recommendations and will ensure that the intent of the TS requirement is met. Therefore, the proposed change will provide improved assurance that the EDGs will be able to meet their safety function.

The proposed change to TS 3.8.5 is an editorial correction and there will be no actual changes to plant design or operation.

Therefore, operation of the facility in accordance with the proposed amendment would 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 Previously Evaluated.

As described above, the proposed change to TS 3.8.3 provides improved assurance that the EDGs will be able to meet their safety function. No new failure modes are introduced. Therefore, no new accident initiators or precursors are introduced by the proposed change.

The proposed change to TS 3.8.5 is an editorial correction and there will be no actual changes to plant design or operation.

Therefore, operation of the facility in accordance with the proposed amendment would 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.

As described above, the proposed change to TS 3.8.3 provides improved assurance that the EDGs will be able to meet their safety function of mitigating events that involve a loss of offsite power. Therefore, the proposed change will preserve any margin of safety.

The proposed change to TS 3.8.5 is an editorial correction and there will be no actual changes to plant design or operation.

Therefore, operation of the facility in accordance with the proposed amendment would 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: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Start Printed Page 21923Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Branch Chief: Douglas A. Broaddus.

Northern States Power Company—Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

Date of amendment request: February 7, 2011.

Description of amendment request: The licensee proposed to amend the MNGP Technical Specifications (TS), revising Surveillance Requirement 3.5.1.7 regarding the Emergency Core Cooling System (ECCS) core spray flow from a minimum of 2800 gpm to a minimum of 2835 gpm. The licensee considers the current minimum flow rate requirement as non-conservative.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration (NSHC) analysis. The NRC staff reviewed the licensee's NSHC analysis and has prepared its own as follows:

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

Response: No.

The purpose of the minimum core spray flow rate requirement is to ensure that the ECCS will perform as designed. Of the postulated accidents and transients previously analyzed in the MNGP Updated Safety Analysis Report, none of them were postulated to be initiated by the ECCS performing as designed.

Furthermore, the consequences of the previously analyzed accidents were not postulated to be exacerbated by the ECCS performing as designed. Accordingly, the probability of occurrence and the consequences of the previously analyzed accidents would not be affected in any way by the proposed amendment to the TS.

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

Response: No.

The proposed amendment does not involve any physical alteration of the plant (no new or different type of equipment will be installed) nor does it change methods and procedures governing plant operation. The proposed amendment will not impose any new or eliminate any old requirements. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed amendment will not have any effect on previously used safety analysis methods, scenarios, acceptance criteria, or assumptions. 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 its own analysis, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the proposed amendment involves no significant hazards consideration.

Attorney for the licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: Robert J. Pascarelli.

Virginia Electric and Power Company, Docket No. 50-281, Surry Power Station, Unit 2, Surry County, Virginia

Date of amendment request: December 16, 2010.

Description of amendment request: This amendment request proposes to revise Technical Specification (TS) 6.4.Q, “Steam Generator (SG) Program,” to exclude portions of the SG tube below the top of the SG tubesheet from periodic tube inspections for Unit 2 during Refueling Outage 23 and the subsequent operating cycle. This amendment request also proposes to revise TS 6.6.A.3, “Steam Generator Tube Inspection Report,” to provide reporting requirements specific to Unit 2 for the temporary alternate repair criteria.

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 previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed change that alters the steam generator inspection/repair criteria and the steam generator inspection reporting criteria does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident.

Of the applicable accidents previously evaluated, the limiting transients with consideration to the proposed change to the steam generator tube inspection and repair criteria are the steam generator tube rupture (SGTR) event and the steam line break (SLB) postulated accidents.

During the SGTR event, the required structural integrity margins of the steam generator tubes and the tube-to-tubesheet joint over the H* distance will be maintained. Tube rupture in tubes with cracks within the tubesheet is precluded by the constraint provided by the tube-to-tubesheet joint. This constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet, and from the differential pressure between the primary and secondary side. Based on this design, the structural margins against burst, as discussed in Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator Tubes,” are maintained for both normal and postulated accident conditions.

The proposed change has no impact on the structural or leakage integrity of the portion of the tube outside of the tubesheet. The proposed change maintains structural integrity of the steam generator tubes and does not affect other systems, structures, components, or operational features. Therefore, the proposed change results in no significant increase in the probability of the occurrence of a SGTR accident.

At normal operating pressures, leakage from primary water stress corrosion cracking below the proposed limited inspection depth is limited by both the tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint. Consequently, negligible normal operating leakage is expected from cracks within the tubesheet region. The consequences of an SGTR event are affected by the primary to secondary leakage flow during the event. However, primary to secondary leakage flow through a postulated broken tube is not affected by the proposed changes since the tubesheet enhances the tube integrity in the region of the hydraulic expansion by precluding tube deformation beyond its initial hydraulically expanded outside diameter. Therefore, the proposed changes do not result in a significant increase in the consequences of a SGTR.

The consequences of a steam line break (SLB) are also not significantly affected by the proposed changes. During a SLB accident, the reduction in pressure above the tubesheet on the shell side of the steam generator creates an axially uniformly distributed load on the tubesheet due to the reactor coolant system pressure on the underside of the tubesheet. The resulting bending action constrains the tubes in the tubesheet thereby restricting primary to secondary leakage below the midplane.

Primary to secondary leakage from tube degradation in the tubesheet area during the limiting accident (i.e., a SLB) is limited by flow restrictions. These restrictions result from the crack and tube-to-tubesheet contact pressures that provide a restricted leakage path above the indications and also limit the degree of potential crack face opening as compared to free span indications.

The probability of a SLB is unaffected by the potential failure of a steam generator tube as the failure of the tube is not an initiator for a SLB event.

The leakage factor of 2.03 is a bounding value for all SGs, both hot and cold legs, in Start Printed Page 21924Table 9-7 of WCAP-17092-P. Also as shown in Table 9-7 of WCAP-17092-P, for Surry for a postulated SLB, a leakage factor of 1.80 has been calculated. However, for Surry, a more conservative leakage factor of 2.03 will be applied to the normal operating leakage associated with the tubesheet expansion region in the condition monitoring (CM) assessment and the operational assessment (OA). Specifically, for the CM assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.03 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the OA, the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.03 and compared to the observed operational leakage.

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?

Response: No.

The proposed change that alters the steam generator inspection/repair criteria and the steam generator inspection reporting criteria does not introduce any new equipment, create new failure modes for existing equipment, or create any new limiting single failures. Plant operation will not be altered, and all safety functions will continue to perform as previously assumed in accident analyses.

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 change involve a significant reduction in a margin of safety?

Response: No.

The proposed change that alters the steam generator inspection/repair criteria and the steam generator inspection reporting criteria maintains the required structural margins of the steam generator tubes for both normal and accident conditions. NEI [Nuclear Energy Institute] 97-06, Revision 2, “Steam Generator Program Guidelines,” and RG 1.121 are used as the bases in the development of the limited tubesheet inspection depth methodology for determining that steam generator tube integrity considerations are maintained within acceptable limits. RG 1.121 describes a method acceptable to the NRC for meeting GDC [General Design Criteria] 14, “Reactor Coolant Pressure Boundary,” GDC 15, “Reactor Coolant System Design,” GDC 31, “Fracture Prevention of Reactor Coolant Pressure Boundary,” and GDC 32, “Inspection of Reactor Coolant Pressure Boundary,” by reducing the probability and consequences of a SGTR. RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation the probability and consequences of a SGTR are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.

For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially oriented cracking, the H* analysis, documented in Section 4 of the license amendment request, defines a length of degradation free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary to secondary leakage during all plant conditions. The methodology for determining leakage provides for large margins between calculated and actual leakage values in the proposed limited tubesheet inspection depth criteria.

Therefore, the proposed change does not involve a significant reduction in any 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, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.

NRC Branch Chief: Gloria Kulesa.

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.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-449, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: May 18, 2010, as supplemented by letter dated March 1, 2011.

Brief description of amendment request: The proposed amendment would revise Technical Specification (TS) 6.8.3.I, “Containment Post-Tensioning System Surveillance Program.” TS 6.8.3.I states that the containment post-tensioning system surveillance program shall be in accordance with American Society of Mechanical Engineers (ASME) Code, Section XI, Subsection IWL, 1992 Edition with 1992 Addenda, as supplemented by 10 CFR 50.55a(b)(2)(viii).

The proposed amendment removes the specific year of the applicable Code edition consistent with Revision 3.1 of NUREG-1431, “Standard Technical Specifications, Westinghouse Plants” and will allow for future updates to the surveillance program when the applicable code edition changes without requiring additional TS changes.

Date of publication of individual notice in the Federal Register: March 22, 2011 (76 FR 16012).

Expiration date of individual notice: April 21, 2011 (public comments); May 23, 2011 (hearing requests).

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 Start Printed Page 21925Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible from the Agencywide Documents Access and Management System (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr.resource@nrc.gov.

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

Date of application for amendments: March 31, 2010, as supplemented by letter dated November 30, 2010.

Brief description of amendments: The amendments revised the Technical Specifications to relocate specific surveillance frequencies to a licensee-controlled program using a risk-informed justification.

Date of issuance: March 29, 2011.

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

Amendment Nos.: 263, 259.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the licenses and the technical specifications.

Date of initial notice in the Federal Register: November 16, 2010 (75 FR 70034). The supplement dated November 30, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 29, 2011.

No significant hazards consideration comments received: No.

Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: March 24, 2010, as supplemented by letters dated November 18, 2010, and March 2, 2011.

Brief description of amendments: The amendments revised the Technical Specifications to relocate specific surveillance frequencies to a licensee-controlled program using a risk-informed justification.

Date of issuance: March 29, 2011.

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

Amendment Nos.: 261, 241.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and the technical specifications.

Date of initial notice in the Federal Register: November 16, 2010 (75 FR 70035).

The supplements dated November 18, 2010, and March 2, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 29, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 15, 2010, as supplemented by letters dated August 30, 2010, September 21, 2010, January 31, 2011, and February 18, 2011.

Brief description of amendment: This amendment request would modify the Technical Specifications to revise the setpoint and setpoint tolerances for safety relief valves (SRVs) and spring safety valves (SSVs) and support the plant modifications associated with the replacement of (1) four Target Rock two-stage SRVs with three-stage SRVs, and (2) two existing Dresser 3.749 inch throat diameter SSVs with Dresser 4.956 inch throat diameter SSVs.

Date of issuance: March 28, 2011.

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

Amendment No.: 235.

Facility Operating License No. DPR-35: The amendment revised the License and Technical Specifications.

Date of initial notice in the Federal Register: May 4, 2010 (75 FR 23812).

The supplemental letters dated August 30, 2010, September 21, 2010, January 31, 2011, and February 18, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated March 28, 2011.

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: August 31, 2010.

Brief description of amendments: The amendments involve administrative changes to the Technical Specifications (TSs). The changes involve: (1) Making an editorial change to Limerick Generating Station (LGS) Unit 1 TS Limiting Condition for Operation (LCO) 3.3.1, Action b; (2) making an editorial change to LGS Units 1 and 2 TS Table 3.3.1-1, Actions 2 and 9; (3) making the layout and format of LGS Unit 1 TS LCO 3.6.5.3 Action requirements consistent with the LGS Unit 2 LCO Action requirements for the same TS; and (4) adding a reference to the minimum required number of operable main turbine bypass valves and the turbine bypass system response time to the core operating limits documented in the Core Operating Limits Report as specified in LGS, Units 1 and 2, TS 6.9.1.9.

Date of issuance: March 31, 2011.

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

Amendment Nos.: 200 and 161.

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

Date of initial notice in the Federal Register: November 2, 2010 (75 FR 67402).

The Commission's related evaluation of the amendment is contained in Safety Evaluation dated March 31, 2011.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating Station, Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: December 15, 2010, as supplemented on February 17, 2011, and March 17, 2011.

Brief description of amendment: The changes revise the Technical Specification (TS) relating to the Safety Limit Minimum Critical Power Ratios (SLMCPRs). The changes result from a cycle specific analysis performed to support the operation of Limerick Start Printed Page 21926Generating Station, Unit 2, in the upcoming Cycle 12. Specifically, the TS changes will revise the SLMCPRs contained in TS 2.1 for two recirculation loop operation and single recirculation loop operation to reflect the changes in the cycle specific analysis. The new SLMCPRs are calculated using Nuclear Regulatory Commission-approved methodology described in NEDE 24011-P-A, General Electric Standard Application for Reactor Fuel, Revision 17.

Date of issuance: April 5, 2011.

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

Amendment No.: 162.

Facility Operating License No. NPF-85. The amendment revised the license and the technical specifications.

Date of initial notice in the Federal Register: February 1, 2011 (76 FR 5620). The supplements dated February 17, 2011, and March 17, 2011, clarified the application, did not expand the scope of the application as originally noticed, and did not change the initial proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

Indiana Michigan Power Company (IandM), Docket No. 50-316, Donald C. Cook Nuclear Plant, Unit 2, Berrien County, Michigan

Date of application for amendment: March 19, 2009, as supplemented on November 20, 2009, February 24, March 11, and March 25, 2011.

Brief description of amendment: The amendment adopts a new analysis of a large-break loss-of-coolant accident, and revises the Technical Specifications to reflect this new analysis, which was performed using a plant-specific adaptation of the NRC-approved methodology set forth in Westinghouse Topical Report WCAP-16009-P-A, “Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM).”

Date of issuance: March 31, 2011.

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

Amendment No.: 297.

Facility Operating License No. DPR-74: Amendment revised the Renewed Operating License and Technical Specifications.

Date of initial notice in the Federal Register: August 11, 2009 (74 FR 40238).

The supplemental information contained clarifying information, did not change the scope of the license amendment request, did not change the NRC staff's initial proposed finding of no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 31, 2011.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 1, 2010.

Brief description of amendments: The amendments revised the inspection scope and repair requirements in Technical Specification (TS) 5.5.9, “Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program,” to exclude portions of the Comanche Peak Nuclear Power Plant (CPNPP), Unit 2, Model D5 SG tubes below the top of the SG tubesheet from periodic SG tube inspections. In addition, the amendments revised TS 5.6.9, “Unit 1 Model D76 and Unit 2 Model D5 Steam Generator Tube Inspection Reports,” to provide reporting requirements specific to CPNPP, Unit 2, for the temporary alternate repair criteria. The changes are applicable only to CPNPP, Unit 2, during Refueling Outage 12 and the subsequent operating cycle.

Date of issuance: April 6, 2011.

Effective date: As of the date of issuance and shall be implemented prior to Mode 4 entry during startup from Unit 2 Refueling Outage 12.

Amendment Nos.: Unit 1—154; Unit 2—154.

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

Date of initial notice in the Federal Register: February 1, 2011 (76 FR 5622).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 6, 2011. No significant hazards consideration comments received: No.

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

Date of application for amendments: April 7, 2009, as supplemented by letters dated June 17 (two letters), September 11, September 25, October 9, November 20 (two letters), November 21 (two letters), November 30, December 8, and December 16 of 2009; and January 7, January 8, January 22, February 11, February 25, March 3, April 15, April 22, April 28, July 8, July 28, August 2, August 9, August 24, October 15, November 1, November 12 (two letters), November 30, and December 21 of 2010. The proposed changes were originally included as part of the April 7, 2009, extended power uprate (EPU) license amendment request, but subsequently divided into a separate licensing action for independent technical review.

Brief description of amendments: The amendment changes the AFW system design and Technical Specifications (TS) 3.7.5, “Auxiliary Feedwater (AFW),” and TS 3.7.6, “Condensate Storage Tank (CST),” resulting from (1) modifications to the AFW system to support requirements for transients and other accidents at EPU conditions; (2) installation of main feedwater isolation valves to support accident mitigation by ensuring that containment pressure does not exceed safety analysis limits; (3) automatic AFW switchover from a CST suction source to a safety-related Service Water source; and (4) instrumentation setpoint changes supporting the aforementioned physical modifications. The upgrades and modifications to the AFW system are being installed to provide additional capacity and reliability for the system. Although the proposed changes are also designed to support the requirements for transients and other accidents at EPU conditions, the changes for this amendment have been evaluated using the current licensing basis.

Date of issuance: March 25, 2011.

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

Amendment Nos.: 238, 242.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revise the License, Appendix C, and the Technical Specifications.

Date of initial notice in the Federal Register: September 21, 2010 (75 FR 57525).

The supplemental letters contained clarifying information and did not change the staff's initial proposed finding of no significant hazards consideration.Start Printed Page 21927

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 25, 2011. No significant hazards consideration comments received: No.

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

Date of application for amendments: April 7, 2009, as supplemented by letters dated June 17, September 11, September 25, November 20, November 30, and December 8 of 2009; February 11, February 25, April 22, April 30, July 21, July 28, August 2, and September 28 of 2010.

Brief description of amendments: The amendment changed the Technical Specifications to support (1) modifications to the AFW system; (2) an EPU to increase plant core thermal power from 1,540 megawatts thermal (MWt) to 1,800 MWt; and (3) update non-conservative RPS and ESFAS setpoints not associated with the EPU. The amendment also modified the RPS instrumentation setpoints of TS Table 3.3.1-1 and the ESFAS instrumentation setpoints of TS Table 3.3.2-1. The changes include both EPU and non-EPU related setpoints. The revised TS allowable values have been calculated to account for new analytical limits, instrument uncertainties, and instrument drift. The changes also include the addition of a new column entitled Nominal Trip Setpoint that was added to provide consistency with the TS Table format in NUREG 1431, “Standard Technical Specifications—Westinghouse Plants,” and Technical Specification Task Force (TSTF)-493, Revision 4, “Clarify the Application of Setpoint Methodology for Limiting Safety System Setting (LSSS) Functions.” The RPS and ESFAS instrumentation uncertainty/setpoint calculations have also been revised to eliminate the use of a single-sided reduction factor in the total loop error determination for LSSS setpoints.

Date of issuance: March 25, 2011.

Effective date: As of the date of issuance, and shall be implemented prior to Unit 1 startup from the Fall 2011 refueling outage (Unit 1) and within 180 days (Unit 2).

Amendment Nos.: 239 and 243.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Technical Specifications/License.

Date of initial notice in the Federal Register: September 21, 2010 (75 FR 57524).

The supplemental letters contained clarifying information and did not change the 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 25, 2011. No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of application for amendment: March 29, 2010, as supplemented by letters dated May 28, and September 30, 2010, and two letters dated February 14, 2011.

Brief description of amendment: The amendment modifies the Technical Specifications (TSs) to extend the allowed outage time for the A and B emergency diesel generators from 72 hours to 14 days.

Date of issuance: March 25, 2011.

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

Amendment No.: 188.

Facility Operating License No. NPF-57: The amendment revised the TSs and the License.

Date of initial notice in the Federal Register: June 29, 2010 (75 FR 37476). The letters dated May, 28, and September 30, 2010, and February 14, 2011 (two letters), provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 25, 2011. No significant hazards consideration comments received: No.

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

Date of amendment request: April 13, 2010, as supplemented by letters dated October 13 and December 21, 2010, and January 18, 2011.

Brief description of amendment: The amendment revised Technical Specification (TS) Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation,” by adding a footnote to Function 8.a concerning the reactor trip P-4 engineered safety feature actuation system interlock. The footnote specifies which functions of the interlock are necessary in each mode in order to meet the limiting condition for operation. Specifically, the functions of tripping the main turbine and isolating main feedwater with a coincident low average temperature would no longer be applicable in MODE 3, which is hot standby. Revised TS Table 3.3.2-1 also identifies that the function of the P-4 interlock that allows arming of the steam dump valves and transfers the steam dump load rejection (Tavg) controller to the plant trip controller is not required in any mode.

Date of issuance: March 30, 2011.

Effective date: The amendment will be effective upon issuance and will be implemented within 90 days from the date of issuance.

Amendment No.: 194.

Renewed Facility Operating License No. NPF-42. The amendment revised the Operating License and Technical Specifications.

Date of initial notice in the Federal Register: June 15, 2010 (75 FR 33844). The supplemental letters dated October 13 and December 21, 2010, and January 18, 2011, 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 March 30, 2011. No significant hazards consideration comments received: No.

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

Date of amendment request: November 30, 2010.

Brief description of amendment: The amendment revised Technical Specification (TS) 5.5.9, “Steam Generator (SG) Program,” to exclude portions of the tube below the top of the steam generator tubesheet from periodic steam generator tube inspections during Refueling Outage 18 and the subsequent operating cycle. In addition, TS 5.6.10, “Steam Generator Tube Inspection Report” will be revised to remove a reference to the previous interim alternate repair criteria and to provide reporting requirements specific to the temporary alternate repair criteria.

Date of issuance: April 6, 2011.

Effective date: The amendment is effective upon issuance and will be implemented prior to MODE 4 entry during startup from Refueling Outage 18.

Amendment No.: 195.

Renewed Facility Operating License No. NPF-42. The amendment revised Start Printed Page 21928the Operating License and Technical Specifications.

Date of initial notice in the Federal Register: February 1, 2011 (76 FR 5623).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 6, 2011. No significant hazards consideration comments received: No.

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Dated at Rockville, Maryland, this 8th day of April 2011.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

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[FR Doc. 2011-9177 Filed 4-18-11; 8:45 am]

BILLING CODE 7590-01-P