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Daily Rules, Proposed Rules, and Notices of the Federal Government

NUCLEAR REGULATORY COMMISSION

[NRC-2012-0205]

Biweekly Notice;

Applications and Amendments to Facility Operating Licenses and Combined 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 or combined license, as applicable, 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 August 8, 2012, to August 21, 2012. The last biweekly notice was published on August 21, 2012, (77 FR 50534).

ADDRESSES: You may submit comments by any of the following methods:

*Federal rulemaking Web site:Go tohttp://www.regulations.govand search for Docket ID NRC-2012-0205. Address questions about NRC dockets to Carol Gallagher; telephone: 301-492-3668; email:Carol.Gallagher@nrc.gov.

*Mail comments to:Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

*Fax comments to:RADB at 301-492-3446.

For additional direction on accessing information and submitting comments, see "Accessing Information and Submitting Comments" in theSUPPLEMENTARY INFORMATIONsection of this document. SUPPLEMENTARY INFORMATION: I. Accessing Information and Submitting Comments A. Accessing Information

Please refer to Docket ID NRC-2012-0205 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and is publicly available, by the following methods:

Federal Rulemaking Web Site:Go tohttp://www.regulations.govand search for Docket ID NRC-2012-0205.

NRC's Agencywide Documents Access and Management System (ADAMS):You may access publicly available documents online in the NRC Library athttp://www.nrc.gov/reading-rm/adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email topdr.resource@nrc.gov. Documents may be viewed in ADAMS by performing a search on the document date and docket number.

NRC's PDR:You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, OneWhite Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2012-0205 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

The NRC cautions you not to include identifying or contact information in comment submissions that you do not want to be publicly disclosed. The NRC posts all comment submissions athttp://www.regulations.govas well as entering the comment submissions into ADAMS, and the NRC does not edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information in their comment submissions that they do not want to be publicly disclosed. Your request should state that the NRC will not edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined 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 section 50.92 of Title 10 of the Code of Federal Regulations (10 CFR), 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 theFederal Registera 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.

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 or combined 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 NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC regulations are accessible electronically from the NRC Library on the NRC's Web site athttp://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 heldwould take place before the issuance of any amendment.

All documents filed in the 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 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email athearing.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 the NRC's public Web site athttp://www.nrc.gov/site-help/e-submittals/apply-certificates.html.System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site athttp://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's 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 athttp://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's public Web site athttp://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 email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's 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's Web site athttp://www.nrc.gov/site-help/e-submittals.html, by email atMSHD.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 the NRC's electronic hearing docket which is available to the public athttp://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 NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC areaccessible electronically through ADAMS in the NRC Library athttp://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's PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email topdr.resource@nrc.gov.

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

Date of amendments request:July 2, 2012.

Description of amendments request:The amendment would revise Technical Specification (TS) 5.5.16 by increasing the calculated peak containment internal pressure (Pa) from 49.4 pounds per square inch gauge (psig) to 49.7 psig for the design basis loss-of-coolant accident (LOCA). In support of the revised Pa, the amendment would also revise the initial internal containment pressure limit in TS 3.6.4 by decreasing the upper bound initial pressure limit from 1.8 psig to 1.0 psig.

Basis for proposed no significant hazards consideration determination:As required by10 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 to Paand the initial containment pressure limit does not alter the assumed initiators to any analyzed event. The probability of an accident previously evaluated will not be increased by this proposed change. The change in Paand the initial containment pressure limit will not affect radiological dose consequence analyses. The radiological dose consequence analyses assume a certain containment atmosphere leak rate based on the maximum allowable containment leakage rate, which is not affected by the change in Pa. The Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J containment leak rate testing program will continue to ensure that containment leakage remains within the leakage assumed in the offsite dose consequence analyses.

Therefore, operation of the facility in accordance with the proposed change to TSs 3.6.4 and 5.5.16 will 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 type of accident from any accident previously evaluated?

Response:No.

The proposed change provides a higher Pathan currently described in the TS. This change is a result of an increase in the mass and energy release input for the LOCA containment response analysis. The Paremains below the containment design pressure of 50 psig because of the change in the initial containment pressure limit, which is an initial condition of the peak pressure calculation. This change does not involve any alteration in the plant configuration, no new or different type of equipment will be installed, or make changes in the methods governing normal plant operation.

Therefore, operation of the facility in accordance with the proposed change to TSs 3.6.4 and 5.5.16 would not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response:No.

The Paremains below the containment design pressure of 50 psig. Since the radiological consequence analyses are based on the maximum allowable containment leakage rate, which is not being revised, the change in the calculated peak containment pressure does not represent a significant change in the margin of safety.

Therefore, operation of the facility in accordance with the proposed change to TSs 3.6.4 and 5.5.16 does not involve a significant reduction in the margin of safety.

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

Attorney for licensee:Steven L. Miller, General Counsel, Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200c, Baltimore, MD 21202.

NRC Branch Chief:George Wilson.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit 2,New London County, Connecticut

Date of amendment request:July 31, 2012.

Description of amendment request:The proposed amendment would revise the Millstone Power Station, Unit 2 (MPS2) Technical Specification requirements regarding steam generator tube inspections and reporting as described in TSTF-510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection;” however, Dominion Nuclear Connecticut, Inc. is proposing minor variations and deviations from TSTF-510.

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 revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. The proposed change to reporting requirements and clarifications of the existing requirements have no affect on the probability or consequences of SGTR.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response:No.

The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

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

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

Response:No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of thephysical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

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

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

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

NRC Branch Chief:George A. Wilson.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit 3, New London County, Connecticut

Date of amendment request:July 31, 2012.

Description of amendment request:The proposed amendment would revise the Millstone Power Station, Unit 3 (MPS3) Technical Specification requirements regarding steam generator tube inspections and reporting as described in TSTF-510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection;” however, Dominion Nuclear Connecticut, Inc. is proposing minor variations and deviations from TSTF-510.

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 revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. The proposed change to reporting requirements and clarifications of the existing requirements have no affect on the probability or consequences of SGTR.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response:No.

The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

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

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

Response:No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

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

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

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

NRC Branch Chief:George A. Wilson.

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

Date of amendment request:June 6, 2012.

Description of amendment request:The proposed amendment would modify Braidwood and Byron Technical Specifications (TS) to add a Note to Surveillance Requirements (SR) 3.3.1.7, 3.3.1.8, and 3.3.1.12 in TS 3.3.1, “Reactor Trip System (RTS) Instrumentation,” and SRs 3.3.2.2 and 3.3.2.6 in TS 3.3.2, “Engineered Safety Features Actuation System (ESFAS) Instrumentation,” to exclude the Solid State Protection System input relays from the Channel Operational Test Surveillance for RTS and ESFAS Functions with installed bypass capability which the U.S. Nuclear Regulatory Commission (NRC) approved by letters dated March 30, 2012, and April 9, 2012.

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 Reactor Protection System (RPS) and ESFAS provide plant protection and are part of the accident mitigating response. The RTS and ESFAS functions do not themselves act at a precursor or an initiator for any transient or design basis accident. Therefore, the proposed change does not significantly increase the probability of any accident previously evaluated.

The proposed change does not alter the design assumptions, conditions, or configuration of the facility. The structural and functional integrity of the RTS and ESFAS, and any other plant system, is unaffected. The proposed change does not alter or prevent the ability of any structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the applicable acceptance criteria. Surveillance testing in the bypass condition will not cause any design or analysis acceptance criteria to be exceeded

The impact of using bypass testing capability upon nuclear safety have been previously evaluated by the NRC and determined to be acceptable in [Westinghouse Atomic Power] WCAP 10271-P-A, Revision 1, WCAP 14333-P-A, Revision 1, and WCAP 15376-P-A, Revision 1. Thus, testing in bypass does not involvea significant increase in the probability or consequences of an accident previously evaluated.

Implementation of the bypass testing capability does not affect the integrity of the fission product barriers utilized for the mitigation of radiological dose consequences as a result of an accident. The plant response as modeled in the safety analyses is unaffected by this change. Hence, the release used as input to the dose calculations are unchanged from those previously assumed.

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

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

Response:No.

The proposed change does not result in a change in the manner in which the RTS and ESFAS provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed change in implemented. In addition, no new failure modes are being created for any plant equipment. The change does not result in the creation of any changes to the existing accident scenarios nor do they create any new or different accident scenarios. The types of accidents defined in the UFSAR [Updated Final Safety Analysis Report] continue to represent the credible spectrum of events to be analyzed which determine safe operation.

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?

Response:No.

No safety analyses are changed or modified as a result of the proposed TS change to reflect installed bypass testing capability. The proposed change does not alter the manner in which the safety limits, limiting safety system setpoints, of limiting conditions for operation are determined. Margins associated with the current applicable safety analyses acceptance criteria are unaffected. The current safety analyses remain bounding since their conclusions are not affected by performing surveillance testing in bypass. The safety systems credited in the safety analyses will continue to be available to perform their mitigation functions.

Redundant RTS and ESFAS trains are maintained, and diversity with regard of the signals that provide reactor trip and engineered safety features actuation is also maintained. All signals credited as primary or secondary, and all operator actions credited in the accident analyses will remain the same. The proposed change will not result in plant operation in a configuration outside the design basis. Although there was no attempt to quantify any positive human factors benefit due to excluding the relays from the [Channel Operational Text] COT Surveillance for those RTS and ESFAS Functions that have installed bypass test capability, it is expected that there would be a new benefit due to a reduced potential for spurious reactor trips and actuations associated with testing.

Implementation of the proposed change is expected to result in an overall improvement of safety, as reduced testing will result in fewer inadvertent reactor trips, less frequent actuation of ESFAF components, less frequent distraction of operations personnel with significant affecting RTS and ESFAS reliability.

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

Based on the above evaluation, EGC concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, (c), and, accordingly, a finding of no significant hazards consideration is justified.

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 requested amendments involve no significant hazards consideration.

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

NRC Branch Chief:Michael I. Dudek.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of application for amendments:July 18, 2012.

Description of amendment request:The proposed amendment would revise the Technical Specifications (TSs) for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 to change the operability requirements for the normal heat sink (NHS). The NHS for PBAPS is the Susquehanna River. Currently, in accordance with TS 3.7.2, the NHS is considered operable with a maximum water temperature of 90 °F. However, TS 3.7.2 also currently contains provisions to allow plant operation to continue if the NHS water temperature exceeds the 90 °F limit. Specifically, the NHS is still considered operable as long as the NHS temperature: (1) does not exceed 92 °F and; (2) is verified at least once per hour to be less than or equal to 90 °F when averaged over the previous 24-hour period. The proposed amendment would change the NHS water temperature limit such that the NHS would be considered operable as long as the maximum water temperature was less than or equal to 92 °F.

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 allows plant operation to continue if the Normal Heat Sink (NHS) temperature does not exceed 92 °F. The water temperature limit imposed for the NHS exists to ensure the ability of safety systems to mitigate the consequences of an accident and does not involve the prevention or identification of any precursors of an accident. The water temperature of the NHS cannot adversely affect the initiator of any accident previously evaluated. This change does not affect the normal operation of the plant to the extent that any accident previously evaluated would be more likely to occur.

The safety objective of the water temperature limit for the NHS is to ensure that the heat removal capability of the Emergency Service Water (ESW) and High Pressure Service Water (HPSW) Systems is adequate to allow safety related equipment that is relied upon to mitigate the consequences of an accident or operational transient to perform its design function. The design basis heat removal capability of the affected components and systems is maintained at the NHS temperature limit, thus ensuring that the affected safety related components continuously perform their safety related function at the NHS temperature limit. The limits for equipment degradation ensure that the affected components continue to perform their design basis function. Consequently, the affected components maintain their design basis capability as previously assumed in [the] plant safety analyses.

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

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 allows plant operation to continue if the Normal Heat Sink (NHS) temperature does not exceed 92 °F. The method of operation of components (heat exchangers, coolers, etc.), which rely on the NHS for cooling, is not altered by this activity. The water temperature limit imposed for the NHS exists to ensure the ability of plant safety equipment to mitigate the consequences of an accident and does not have the potential to create an accident initiator. This activity does not involve a physical change to any plant structure, system or component that is considered an accident initiator. The design basis heat removal capability of the affected components is maintained.

This license amendment request does not involve any changes to the operation, testing, or maintenance of any safety-related, orotherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases.

Therefore, no new failure modes are introduced and the possibility of a new or different kind of accident is not created.

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

Response:No.

Operation of PBAPS, Units 2 and 3 under the NHS temperature limit (92 °F) does not reduce the margin of safety as defined in the basis for any Technical Specification. Technical Specification Surveillance Requirement (SR) 3.7.2.2 defines the value for satisfying the Limiting Condition for Operation for the temperature of the NHS. A portion of the Technical Specification Bases for SR 3.7.2.2 states:

Verification of the Normal Heat Sink temperature ensures that the heat removal capability of the ESW and HPSW Systems is within the DBA [design-basis accident] analysis.

The basis for SR 3.7.2.2 has not changed as a result of the proposed [change]. The heat removal capability of the components that rely on the ESW and HPSW Systems for cooling is based on the Technical Specification temperature limit (92 °F) of the NHS and the performance capability of the equipment. Periodic testing and cleaning are required to verify and ensure that the assumed degree of degradation is not reached. The limits for equipment degradation ensure that affected components continue to perform their design basis function.

Therefore, since the design basis capability of the affected components is maintained at the NHS temperature limit (92 °F), this change does not involve a significant reduction in the margin of safety.

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

Attorney for Licensee:Mr. J. Bradley Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief:Meena K. Khanna.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Units 1 and 2, Beaver County, Pennsylvania

Date of amendment request:July 25, 2012.

Description of amendment request:The proposed amendment would modify Technical Specification (TS) 3.1.3 to allow the normally required near-end of life Moderator Temperature Coefficient (MTC) measurement to not be performed under certain conditions. If these specified conditions are met, the MTC measurement would be replaced by a calculated value.

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, with NRC edits in brackets:

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

Response:No.

This amendment request would change the near-end of life (EOL) moderator temperature coefficient (MTC) surveillance requirement (SR) to allow [] the required MTC measurement [to be eliminated] under certain conditions. This change would not result in physical alteration of a plant structure, system or component, or installation of new or different types of equipment. Modification of the surveillance requirement under certain conditions would not affect the probability of accidents previously evaluated in the Updated Final Safety Analysis Report (UFSAR) or cause a change to any of the dose analyses associated with the UFSAR accidents because accident mitigation functions would remain unchanged. Existing MTC TS limits would remain unchanged and would continue to be satisfied.

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

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

Response:No.

This amendment request would change the near EOL MTC SR to allow [] the required MTC measurement [to be eliminated] under certain conditions. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. No physical plant alterations are made as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety related system. MTC is a variable that must remain within limits but is not an accident initiator.

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?

Response:No.

This amendment request would change the near EOL MTC SR to allow [] the required MTC measurement to be eliminated under certain conditions. The margin of safety associated with the acceptance criteria of accidents previously evaluated in the UFSAR is unchanged. The proposed change would have no affect on the availability, operability, or performance of the safety-related systems and components. A change to a surveillance is proposed based on an alternate method of confirming that the surveillance requirement is met. The Technical Specification limiting condition for operation (LCO) limits for MTC remain unchanged.

The Technical Specifications establish limits for the moderator temperature coefficient based on assumptions in the UFSAR accident analyses. Applying the conditional [elimination of] the moderator temperature coefficient measurement changes the method of meeting the surveillance requirement; however this change does not modify the TS values and ensures adherence to the current TS limits. The basis for derivation of the moderator temperature coefficient limits from the moderator density coefficient assumed in the accident analysis would not change.

Therefore, the margin of safety as defined in the TS is not reduced and the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and based on this review, with the edits noted above, 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 W. Jenkins, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief:Meena Khanna.

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

Date of amendment request:July 16, 2012, as supplemented by letter dated August 10, 2012.

Description of amendment request:The proposed amendments would modify Technical Specification (TS) requirements regarding steam generator tube inspections and reporting as described in Technical Specification Task Force Traveler 510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.” The proposed changes would revise TS 3/4.4.5, “Steam Generator (SG) Tube Integrity,” TS 6.8.4.j, “Steam Generator (SG) Program.” and TS 6.9.1.8, “Steam Generator Tube Inspection Report.”

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 revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrityand SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response:No.

The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

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

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

Response:No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant system pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant system pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

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

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

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

NRC Acting Branch Chief:Jessie F. Quichocho.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant (HNP), Units 1 and 2, Appling County, Georgia

Date of amendment request:July 5, 2012.

Description of amendment request:The proposed amendments would revise Technical Specification (TS) Limiting Condition for Operation (LCO) for the plant service water (PSW) and ultimate heat sink (UHS). Specifically, the surveillance requirement (SR) for the minimum water level in each PSW pump well of the intake structure would be revised from the existing value to a lower value. This change is based on updated design basis analyses that demonstrate that at the new minimum level sufficient water inventory remains available from the Altamaha River for PSW and residual heat removal service water (RHRSW) to handle Loss of Coolant Accident (LOCA) cooling requirements for 30 days post-accident with no additional makeup water source available.

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 TS change revises the minimum water level in the PSW pump well, as required by SR 3.7.2.1, from 60.7 [feet] ft [mean sea level] MSL to 60.5 ft MSL. TS SR 3.7.2.1 verifies that the UHS is OPERABLE by ensuring the water level in the PSW pump well of the intake structure is sufficient for the PSW, RHRSW and standby service water pumps to supply post-LOCA cooling requirements for 30 days. The safety function of the UHS is to mitigate the impact of an accident. The proposed TS change does not result in or require any physical changes to HNP systems, structures, and components, including those intended for the prevention of accidents. The potential impact of the lower PSW pump well minimum water level on pump operation requirements, supply of water for 30 days post-LOCA, and potential environmental impact have been evaluated and found to be acceptable.

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 TS change revises the minimum water level in the PSW pump well, as required by SR 3.7.2.1, from 60.7 ft MSL to 60.5 ft MSL. TS SR 3.7.2.1 verifies that the UHS is OPERABLE by ensuring the water level in the PSW pump well of the intake structure is sufficient for the PSW, RHRSW and standby service water pumps to supply post-LOCA cooling requirements for 30 days. The proposed TS change does not result in or require any physical changes to HNP systems, structures, and components. The potential impact of the lower PSW pump well minimum water level on pump operation requirements, supply of water for 30 days post-LOCA, and potential environmental impact have been evaluated and found to be acceptable.

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

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

Response:No.

The proposed TS change revises the minimum water level in the PSW pump well, as required by SR 3.7.2.1, from 60.7 1t MSL to 60.5 1t MSL. TS SR 3.7.2.1 verifies that the UHS is OPERABLE by ensuring the water level in the PSW pump well of the intake structure is sufficient for the PSW, RHRSW and standby service water pumps to supply post-LOCA cooling requirements for 30 days. The proposed TS change does not result in or require any physical changes to HNP systems, structures, and components. The potential impact of the lower PSW pump well minimum water level on pump operation requirements, supply of water for 30 days post-LOCA, and potential environmental impact have been evaluated and found to be acceptable.

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

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

Attorney for licensee:Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

NRC Branch Chief:Nancy L. Salgado.

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

Date of amendment request:May 2, 2012.

Description of amendment request:The amendment would revise Technical Specification (TS) 3.6.6, “Containment Spray and Cooling Systems,” to replace the 10-year surveillance frequency for testing the containment spray nozzles as required by TS Surveillance Requirement 3.6.6.8 with an event-based frequency.

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 Containment Spray System and its spray nozzles are not accident initiators and therefore, the proposed change does not involve a significant increase in the probability of an accident. The revised surveillance requirement will require event-based Frequency verification in lieu of fixed Frequency verification. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that may initiate an analyzed event. The proposed change will not alter the operation or otherwise increase the failure probability of any plant equipment that can initiate an analyzed accident.

This change does not affect the plant design. There is no increase in the likelihood of formation of significant corrosion products. Due to their location at the top of the containment, introduction of foreign material into the spray headers is unlikely. Foreign material introduced during maintenance activities would be the most likely source for obstruction, and verification following such maintenance would confirm the nozzles remain unobstructed. Since the Containment Spray System will continue to be available to perform its accident mitigation function, the consequences of accidents previously evaluated are not significantly increased.

Therefore, the consequences of an accident previously evaluated are not significantly affected by the proposed change.

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

Response: No.

The proposed change will not physically alter the plant (no new or different type of equipment will be installed) or change the methods governing normal plant operation. The proposed change does not introduce new accident initiators or impact assumptions made in the safety analysis. Testing requirements continue to demonstrate that the Limiting Conditions for Operation are met and the system components are functional.

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

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

Response:No.

The system is not susceptible to corrosion-induced obstruction or obstruction from sources external to the system. Maintenance activities that could introduce foreign material into the system would require subsequent verification to ensure there is no nozzle blockage. The spray header nozzles are expected to remain unblocked and available in the event that the safety function is required. Therefore, the capacity of the system would remain unaffected. The proposed change does not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by this change.

Therefore the