81 Fed. Reg. 21593 (Apr. 12, 2016)

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

Download PDF

Federal RegisterApr 12, 2016

81 Fed. Reg. 21593 (Apr. 12, 2016)

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to 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 March 15, 2016, to March 28, 2016. The last biweekly notice was published on March 29, 2016.

DATES:

Comments must be filed by May 12, 2016. A request for a hearing must be filed by June 13, 2016.

ADDRESSES:

You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0073. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov.
Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT:

Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1927, email: Lynn.Ronewicz@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2016-0073 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0073.
NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://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 to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section of this document.
NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2016-0073, facility name, unit number(s), application date, and subject in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov,, as well as entering the comment submissions into ADAMS. The NRC does not routinely 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 that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 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; (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.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

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 “Agency Rules of Practice and Procedure” 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's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, 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 set forth 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 with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with NRC regulations, policies and procedures.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii).

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, 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 unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by June 13, 2016. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions for leave to intervene set forth in this section, except that under § 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c).

If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by June 13, 2016.

B. Electronic Submissions (E-Filing)

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's 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 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 the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.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 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'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 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's 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 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 NRC'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 public Web site at http://www.nrc.gov/site-help/e-submittals.html,, by email to 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 the 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. However, in some instances, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. 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.

For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, North Carolina; Docket No. 50-261, H. B. Robinson Steam Electric Plant (RNP) Unit No. 2, Darlington County, South Carolina; and Docket No. 50-400, Shearon Harris Nuclear Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina

Date of amendment request: February 1, 2016. A publicly-available version is in ADAMS under Accession No. ML16040A077.

Description of amendment request: The amendments would change the licensee's name from Duke Energy Progress, Inc. to Duke Energy Progress, LLC.

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

Response: No.

The proposed changes do not involve a significant increase in the probability of any accident previously evaluated because no accident initiators or assumptions are affected. The proposed conversion and name change is administrative in nature and has no direct effect on any plant system, plant personnel qualifications, or the operation and maintenance of BSEP, RNP, and HNP.

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

Response: No.

The proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated because the proposed name change is administrative in nature and does not involve new failure mechanisms, malfunctions, or accident initiators. The proposed changes have no direct effect on any plant system, plant personnel qualifications, or operation and maintenance of BSEP, RNP, and HNP.

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

Response: No.

The proposed changes will not involve a significant reduction in the margin of safety because the proposed changes do not involve changes to the initial conditions contributing to accident severity or consequences, or reduce response or mitigation capabilities. The proposed name change is administrative in nature and has no direct effect on any plant system, plant personnel qualifications, or operation and maintenance of BSEP, RNP, and HNP.

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

Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC 28202.

NRC Branch Chief: Benjamin G. Beasley.

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana

Date of amendment request: October 29, 2015. A publicly-available version is in ADAMS under Accession No. ML15307A293.

Description of amendment request: The amendment proposes to modify Technical Specification (TS) 5.5.13, “Primary Containment Leakage Rate Testing Program,” by incorporating Nuclear Energy Institute (NEI) topical report 94-01, Revision 3-A, as the implementation document for the RBS performance-based containment leakage rate testing program. Based on the guidance in NEI 94-01, Revision 3-A, the proposed change would allow the RBS Type A Test (Integrated Leak Rate Test) frequency to be extended from 10 to 15 years, and the Type C Tests (Local Leak Rate Tests) frequency to be extended from 60 to 75 months. Additionally, the amendment proposes to modify Surveillance Requirement (SR) 3.6.5.1.3 to extend the frequency of the Drywell Bypass Test from 10 to 15 years and to revise its allowed extension per SR 3.0.2 from 12 to 9 months.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below:

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

Response: No.

The proposed amendment incorporates NEI topical report 94-01, Revision 3-A, into TS 5.5.13 as the basis for the RBS containment leakage rate testing program, which would allow for extensions to the frequencies of the Type A and Type C Tests. The proposed amendment also requests an extension to the Drywell Bypass Test frequency. The proposed changes do not involve any physical changes to the plant or any changes in the normal operation or control of the plant. In its license amendment request, the licensee identified the loss-of-coolant accident (LOCA) inside containment and the fuel handling accident (FHA) as the previously evaluated accidents in the Updated Safety Analysis Report that could potentially be impacted by the change. Changing the frequency of containment leakage rate testing has no impact upon the likelihood of a LOCA or of an FHA. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

The guidelines in NEI 94-01, Revision 3-A, provide a framework for a licensee's containment leakage rate testing program, the purpose of which is to ensure that the primary containment limits the uncontrolled release of radioactivity to the environment during a design-basis accident. As part of its amendment request, the licensee evaluated the potential consequences of extending the test intervals and determined that the change in risk was estimated to be acceptably small and within the guidelines, as published in Regulatory Guide 1.174. The proposed amendment does not change the overall containment leakage rate limit specified by the TSs. Therefore, it is concluded that the proposed amendment does not significantly increase the consequences of an accident previously evaluated.

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

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

Response: No.

The proposed changes do not involve any physical changes to the plant or any changes in the normal operation or control of the plant. The proposed changes do not create any new accident precursors or initiators, and do not change any existing accident precursors or initiators, as described in the RBS safety analyses.

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 adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for the development of the RBS performance-based leakage rate testing program, to allow for frequency extensions for the Type A and Type C Tests. The proposed amendment also requests an extension to the Drywell Bypass Test frequency. The proposed changes do not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the containment leakage rate testing program, as defined in the TSs, ensure that the primary containment will continue to provide a leaktight barrier to the uncontrolled release of radioactivity to the environment during a design-basis accident. The proposed amendment does not change the overall containment leakage rate limit specified by the TSs. Additionally, the proposed amendment does not include any changes to the Containment Inservice Inspection Plan at RBS, which serves to provide a high degree of assurance that the containment will not degrade in a manner that is not detectable by the Type A Test.

Based on the above discussion, 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 review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Meena K. Khanna.

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

Date of amendment request: January 14, 2016. A publicly available version is in ADAMS under Accession No. ML16021A459.

Description of amendment request: The amendment would revise the PNPS Emergency Plan to decrease the Emergency Response Organization (ERO) staff training requirements identified for the “on-site” Chemistry Technician.

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, along with NRC edits in square brackets:

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

Response: No.

The proposed training requirements change has no effect on normal plant operation or on any accident initiator. The change affects the response to radiological emergencies addressed in the SEP [site emergency plan]. The ability of the emergency response organization to respond adequately to radiological emergencies has been evaluated. Changes in the training provided to the on-shift organization, such as the reassignment of key on-shift emergency personnel to perform related RP [radiation protection] functions, provide assurance of an effective emergency response without competing or conflicting duties. An analysis was also performed on the effect of the proposed change on the timeliness of performing major tasks for the major functional areas of the SEP. The analysis concluded that the reduction in training requirements for the “on-shift” Chemistry Technician to support the initial RP support tasks does not affect the ability to perform the required RP Technician or Chemistry Technician tasks.

Therefore, the change in ERO staff training does not increase 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 affects the training requirements for the “on-shift” Chemistry Technician and for supplementing onsite personnel in response to a radiological emergency. It has been evaluated and determined not to significantly affect the ability to perform required or related functions. It has no effect on the plant design or on the normal operation of the plant and does not affect how the plant is physically operated under emergency conditions. The reduction in ERO training requirements for the “on shift” Chemistry Technician in the SEP does not affect the plant operating procedures which are performed by plant staff during all plant conditions.

No new or different accidents are postulated to occur and there are no changes in any of the accidents previously evaluated.

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

Response: No.

The proposed change does not affect plant design or method of operation. 10 CFR 50.47(b) and 10 CFR 50 Appendix E establish emergency planning standards and requirements that require adequate staffing, satisfactory performance of key functional areas and critical tasks, and timely augmentation of the response capability. Since the SEP was originally developed, there have been improvements in the technology used to support the SEP functions and in the capabilities of onsite personnel. A functional analysis was performed on the effect of the proposed change on the timeliness of performing major tasks for the functional areas of the SEP. The analysis concluded that a reduction in training requirements for the “on-shift” Chemistry Technician would not significantly affect the ability to perform the required SEP tasks. Thus, the proposed change has been determined not to adversely affect the ability to meet the emergency planning standards as described in 10 CFR 50.47(b) and requirements in 10 CFR 50 Appendix E.

The proposed ERO staff training change does not involve a reduction in any margin of safety. The proposed change is consistent with the original and current ERO staffing levels implemented at PNPS.

Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Travis L. Tate.

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

Date of amendment request: February 4, 2016. A publicly available version is in ADAMS under Accession No. ML16035A227.

Description of amendment request: The amendments would add Surveillance Requirement (SR) 3.5.2.10 to the list of applicable SRs shown in SR 3.5.3.1.

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 staff corrections shown in [brackets]:

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

Response: No.

The proposed LAR [license amendment request] is purely an administrative change; therefore, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the TS [technical specifications] for which SR 3.5.2.10 is applicable, continue to be operable and capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an[y] 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 LAR is purely an administrative change. The proposed change to add SR 3.5.2.10 to the list of applicable surveillances in SR 3.5.3.1 does not create a new or different kind of accident [than] previously evaluated.

The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the change does not impose any new or different requirements. The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

The proposed LAR is purely an administrative change to add SR 3.5.2.10 to the list of applicable surveillances in SR 3.5.3.1.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the Final Safety Analysis Report and Bases to TS). Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis.

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

Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 60555.

NRC Branch Chief: Travis L. Tate.

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

Date of amendment request: January 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16029A418.

Description of amendment request: The amendment would revise the post-loss-of-coolant-accident (post-LOCA) drawdown time for secondary containment from 12 to 19 minutes as described in the CPS Updated Safety Analysis Report and technical specification bases.

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

Response: No.

The proposed change results in additional heat added to Secondary Containment and the resultant increase in the time to achieve and maintain the required negative pressure in Secondary Containment following a LOCA. Neither the additional heat load from DCS [dry-cask storage] activities, nor the resultant increase in the time to achieve and maintain the required negative pressure in Secondary Containment affect any initiator or precursor of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The proposed change results in an increase in the post-LOCA radiological dose to a Control Room occupant. However, the resultant post-LOCA Control Room dose remains within the regulatory limits of 10 CFR 50.67 and GDC [General Design Criterion] 19. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

In summary, 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 alter the design function or operation of Secondary Containment or the Standby Gas Treatment system [SGTS], or the ability of each to perform its design function. EGC has evaluated the post-LOCA pressure response of Secondary Containment assuming the higher heat load, utilizing the design basis short-term pressure response analysis. The results of this analysis validated that SGTS will achieve and maintain the required negative pressure in Secondary Containment within the specified timeframe. The proposed change does not alter the safety limits, or safety analysis associated with the operation of the plant. Accordingly, the change does not introduce any new accident initiators. Rather, this proposed change is the result of an evaluation of the Control Room doses following the most limiting LOCA that can occur at CPS. The proposed change does not introduce any new modes of plant operation. As a result, no new failure modes are introduced.

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.

The revised post-LOCA dose consequences to a Control Room occupant were calculated in accordance with the requirements of 10 CFR 50.67, Regulatory Guide 1.183, and SRP [Standard Review Plan] 15.0.1 and are consistent with the post-LOCA dose calculations approved by the NRC in Amendment No. 167 to the CPS Facility Operating License NPF-62.

The margin of safety is considered to be that provided by meeting the applicable regulatory limits. The additional heat load that is added to Secondary Containment during DCS activities, leading to an increase in Secondary Containment drawdown time results in an increase in Control Room dose following the LOCA design basis accident. However, since the Control Room dose following the design basis accident remains within the regulatory limits, there is not a significant reduction in a margin of safety.

Therefore, operation of CPS in accordance with the proposed change will not involve a significant reduction in a margin of safety.

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

Acting NRC Branch Chief: Justin C. Poole.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-346, Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa County, Ohio

Date of amendment request: December 16, 2015, as supplemented by letters dated February 2 and March 7, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML15350A314, ML16033A085, and ML16067A195.

Description of amendment request: The amendment would allow the licensee to transition the current fire protection program at DBNPS to a performance-based, risk-informed fire protection program consistent with 10 CFR, Section 50.48(c), “National Fire Protection Association Standard NFPA 805.” The 2001 Edition of NFPA 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,” is incorporated by reference into 10 CFR 50.48(c), with exceptions, modifications, and supplementation. The amendment would also allow the licensee to make changes to the DBNPS fire protection program without prior NRC approval, provided that specified conditions are met. The proposed amendment would change the facility operating license, technical specifications, and design basis.

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

Response: No.

Operation of DBNPS in accordance with the proposed amendment does not increase the probability or consequences of accidents previously evaluated. The Updated Final Safety Analysis Report (UFSAR) documents the analyses of design basis accidents (DBAs) at DBNPS. The proposed amendment does not affect accident initiators, nor does it alter design assumptions, conditions, or configurations of the facility that would increase the probability of accidents previously evaluated. Further, the changes to be made for fire hazard protection and mitigation do not adversely affect the ability of SSCs [structures, systems, and components] to perform their design functions for accident mitigation, nor do they affect the postulated initiators or assumed failure modes for accidents described and evaluated in the UFSAR. SSCs required to shut down the reactor safely and to maintain it in a safe and stable condition will remain capable of performing their design functions.

The purpose of the proposed amendment is to permit DBNPS to adopt a new fire protection licensing basis, which complies with the requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance in [Regulatory Guide] RG 1.205, Revision 1. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection requirements that are an acceptable alternative to the 10 CFR 50, Appendix R required fire protection features (69 FR 33536, June 16, 2004). Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been satisfied.

NFPA 805, taken as a whole, provides an acceptable alternative for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 10 CFR 50, meets the underlying intent of the NRC's existing fire protection regulations and guidance, and provides for DID [defense-in-depth]. The goals, performance objectives, and performance criteria specified in Chapter 1 of the standard ensure that, if there are any increases in CDF [core damage frequency] or risk, the increase will be small and consistent with the intent of the Commission's Safety Goal Policy.

Based on this, the implementation of the proposed amendment does not increase the probability of any accident previously evaluated. Equipment required to mitigate an accident remains capable of performing the assumed function(s). The proposed amendment will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The applicable radiological dose criteria will continue to be met. Therefore, the consequences of any accident previously evaluated are not significantly increased with the implementation of the proposed amendment.

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

Response: No.

Operation of DBNPS in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not alter the requirements or functions for systems required during accident conditions. Implementation of the new fire protection licensing basis that complies with the requirements of 10 CFR 50.48(a) and 10 CFR 50.58(c) and the guidance in RG 1.205, Revision 1, will not result in new or different accidents.

The proposed amendment does not adversely affect accident initiators or alter design assumptions, conditions, or configurations of the facility. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to maintain the plant in a safe and stable condition remain capable of performing their design functions.

The proposed amendment does not introduce new or different accident initiators, nor does it alter design assumptions, conditions, or configurations of the facility. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shutdown the reactor and maintain it in a safe and stable condition remain capable of performing their design functions.

The purpose of the proposed amendment is to permit DBNPS to adopt a new fire protection licensing basis that complies with the requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance in Regulatory Guide 1.205, Revision 1. The NRC considers that NFPA 805 provides an acceptable methodology and appropriate performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R required fire protection features (69 FR [Federal Register] 33536, June 16, 2004).

The requirements of NFPA 805 address only fire protection and the impacts of fire on the plant that have previously been evaluated. Based on this, implementation of the proposed amendment would not create the possibility of a new or different kind of accident from any kind of accident previously evaluated. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety-related system as a result of this amendment. Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created with the implementation of this amendment.

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

Response: No.

Operation of DBNPS in accordance with the proposed amendment does not involve a significant reduction in the margin of safety. The proposed amendment does not alter the manner in that safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed to mitigate accidents in the UFSAR. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and to maintain it in a safe and stable condition remain capable of performing their design functions.

The purpose of the proposed amendment is to permit FENOC to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance in RG 1.205, Revision 1. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50 Appendix R required fire protection features (69 FR 33536, June 16, 2004). Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 do not result in a significant reduction in the margin of safety.

The proposed changes are evaluated to ensure that risk and safety margins are kept within acceptable limits. Therefore, the transition to NFPA 805 does not involve a significant reduction in the margin of safety.

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.

Acting NRC Branch Chief: Justin C. Poole.

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

Date of amendment request: March 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16076A433.

Description of amendment request: The amendment would adopt Technical Specification (TS) Task Force (TSTF) Change Traveler TSTF-535, Revision 0, “Revise Shutdown Margin [SDM] Definition to Address Advanced Fuel Designs.” The SDM (i.e., the amount of reactivity by which the reactor is subcritical), is calculated under the conservative conditions that the reactor is Xenon free, the most reactive control rod is outside the reactor core, and the moderator temperature produces the maximum reactivity. For standard fuel designs, maximum reactivity occurs at a moderator temperature of 68 degrees Fahrenheit (°F), which is reflected in the temperature specified in the TSs. New, advanced boiling water reactor fuel designs can have a higher reactivity at moderator shutdown temperatures above 68 °F. Therefore, the proposed amendment, consistent with TSTF-535, Revision 0, seeks to modify the TSs to require the SDM to be calculated at whatever temperature produces the maximum reactivity (i.e., temperatures at or above 68 °F). The availability of this TS improvement was announced in the Federal Register (FR) published on February 26, 2013 (78 FR 13100), as part of the Consolidated Line Item Improvement Process, and has been requested with no variations or deviations.

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 definition of SDM. SDM is not an initiator to any accident previously evaluated. Accordingly, the proposed change to the definition of SDM has no effect on the probability of any accident previously evaluated. SDM is an assumption in the analysis of some previously evaluated accidents and inadequate SDM could lead to an increase in consequences for those accidents. However, the proposed change revises the SDM definition to ensure that the correct SDM is determined for all fuel types at all times during the fuel cycle. As a result, the proposed change does not adversely affect the consequences of any accident previously evaluated.

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

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

Response: No.

The proposed change revises the definition of SDM. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operations. The change does not alter assumptions made in the safety analysis regarding SDM.

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

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

Response: No.

The proposed change revises the definition of SDM. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change ensures that the SDM assumed in determining safety limits, limiting safety system settings or limiting conditions for operation is correct for all Boiling Water Reactor fuel types at all times during the fuel cycle.

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, P.O. Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Meena K. Khanna.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, Fairfield County, South Carolina

Date of amendment request: December 16, 2015, as supplemented by letter dated March 7, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML15356A048 and ML16069A021, respectively.

Description of amendment request: The licensee proposes to revise TS 3/4.3.1, “Reactor Trip System Instrumentation,” and TS 3/4.3.2, “Engineered Safety Feature Actuation System Instrumentation,” to implement the Allowed Outage Time, Bypass Test Time, and Surveillance Frequency changes approved by the NRC in WCAP-15376-P-A, Rev. 1, “Risk-Informed Assessment of the Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS) Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times.”

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

Response: No.

The overall protection system performance will remain within the bounds of the previously performed accident analyses since no hardware changes are proposed. The same reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) instrumentation will continue to be used. The protection systems will continue to function in a manner consistent with the plant design basis. These changes to the Technical Specifications do not result in a condition where the design, material, and construction standards that were applicable prior to the change are altered.

The proposed changes will not modify any system interfaces. The proposed changes will not affect the probability of any event initiators. There will be no degradation in the performance of or an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Final Safety Analysis Report (FSAR).

The determination that the results of the proposed changes are acceptable was established in the NRC Safety Evaluation prepared for WCAP-1 5376-P-A (issued by letter dated December 20, 2002 [ML023540534]). Implementation of the proposed changes will result in an insignificant risk impact. Applicability of these conclusions has been verified through plant-specific reviews and implementation of the generic analysis results in accordance with the NRC Safety Evaluation conditions.

The proposed changes to the Completion Times, bypass test times, and Surveillance Frequencies reduce the potential for inadvertent reactor trips and spurious engineered safety feature (ESF) actuations, and therefore do not increase the probability of any accident previously evaluated. The proposed changes do not change the response of the plant to any accidents and have an insignificant impact on the reliability of the RTS and ESFAS signals. The RTS and ESFAS instrumentation will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This is demonstrated by showing that the impact on plant safety as measured by the increase in core damage frequency (CDF) is less than 1.0E-06 per year and the increase in large early release frequency (LERF) is less than 1.0E-07 per year. In addition, for the Completion Time changes, the incremental conditional core damage probabilities (ICCDP) and incremental conditional large early release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, respectively. These changes meet the acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and ESFAS instrumentation will continue to perform their functions with high reliability as originally assumed, and the risk impact as measured by the ΔCDF, ΔLERF, ICCDP, and ICLERP risk metrics is within the acceptance criteria of existing regulatory guidance, there will not be a significant increase in the consequences of any accidents.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes are consistent with safety analysis assumptions and resultant consequences.

Therefore, 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.

There are no hardware changes nor are there any changes in the method by which any safety-related plant system performs its safety function. The proposed changes will not affect the normal method of plant operation. No performance requirements will be affected or eliminated.

The proposed changes will not result in physical alteration to any plant system nor will there be any change in the method by which any safety-related plant system performs its safety function. The proposed changes do not include any changes to the instrumentation setpoints or changes to the accident analysis assumptions.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of these changes. There will be no adverse effect or challenges imposed on any safety-related system as a result of these changes.

Therefore, the proposed changes do 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 proposed changes do not affect the acceptance criteria for any analyzed event nor is there a change to any Safety Analysis Limit (SAL). There will be no effect on the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions.

The redundancy of RTS and ESFAS is maintained, and diversity with regard to the signals that provide reactor trip and ESF 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 changes will not result in plant operation in a configuration outside the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in Regulatory Guides 1.174 and 1.177. Although there was no attempt to quantify any positive human factors benefit due to increased Completion Times and bypass test times, it is expected that there would be a net benefit due to a reduced potential for spurious reactor trips and actuations associated with testing.

Implementation of the proposed changes is expected to result in an overall improvement in safety, as follows:

(a) Reduced testing should result in fewer inadvertent reactor trips, less frequent actuation of ESFAS components, less frequent distraction of operations personnel without significantly affecting RTS and ESFAS reliability.

(b) The Completion Time extensions for the reactor trip breakers should provide additional time to complete test and maintenance activities while at power, potentially reducing the number of forced outages related to compliance with reactor trip breaker Completion Times, and provide consistency with the Completion Times for the logic trains.

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

Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina Electric & Gas Company, P.O. Box 764, Columbia, SC 29218.

NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia

Date of amendment request: February 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16046A009.

Description of amendment request: The proposed change would amend Combined License Nos. NPF-91 and NPF-92 for the VEGP Units 3 and 4. The requested amendment proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 information and involves related changes to the associated plant-specific Tier 2* information. Specifically, the proposed departures consist of changes to UFSAR text and tables, and information incorporated by reference into the UFSAR related to updates to WCAP-16096, “Software Program Manual for Common Q^TM Systems,” and WCAP-16097, “Common Qualified Platform Topical Report.”

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

Response: No.

WCAP-16096 (Common Q Software Program Manual) was updated to Revision 4 to reference later NRC endorsed regulatory guides and standards and update the requirements for the software design and development processes for the Common Q portion of the AP1000 Protection and Safety Monitoring System (PMS). WCAP-16097 (Common Q Topical Report) was updated to Revision 3 to describe new Common Q components and standards currently used for the AP1000 PMS implementation of the Common Q platform. These two WCAPs have been reviewed and approved by the NRC in Safety Evaluations dated February 7, 2013. WCAP-15927 was updated to reference the newest revisions of WCAP-16096 and WCAP-16097 and for editorial corrections. The proposed activity adopts the updated versions as incorporated by reference documents into the UFSAR. Other proposed document changes support the implementation of the updated versions of WCAP-16096, WCAP-16097, and WCAP-15927.

The Common Q platform is an acceptable platform for nuclear safety-related applications. The Common Q system meets the requirements of 10 CFR part 50, Appendix A, General Design Criteria (Criteria 1, 2, 4, 13, 19, 20, 21, 22, 23, 24, and 25), the Institute of Electrical and Electronics Engineers Standard 603-1991 for the design of safety-related reactor protection systems, engineered safety features systems and other plant systems, and the guidelines of Regulatory Guide 1.152 and supporting industry standards for the design of digital systems.

Because the Common Q platform and the PMS implementation of the Common Q platform meet the criteria in the applicable General Design Criteria, the revisions to these documents do not affect the prevention and mitigation of abnormal events, such as accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses as described in the licensing basis. The incorporation of the updated documents does not adversely affect the interface with any structure, system, or component accident initiator or initiating sequence of events. Thus, the probabilities of the accidents previously evaluated in the UFSAR are not affected.

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

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

Response: No.

The proposed changes to adopt the updated WCAP-16096, WCAP-16097, and WCAP-15927 into the UFSAR do not adversely affect the design or operation of safety-related equipment or equipment whose failure could initiate an accident beyond what is already described in the licensing basis. These changes do not adversely affect fission product barriers. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change.

Therefore, this activity 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 changes to adopt the updated WCAP-16096, WCAP-16097, and WCAP-15927 into the UFSAR do not adversely affect the design, construction, or operation of any plant SSCs, including any equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. Furthermore, no system function, design function, or equipment qualification will be adversely affected by the changes.

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

Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

Acting NRC Branch Chief: John McKirgan.

Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: March 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16071A333.

Description of amendment request: The amendments would revise the Technical Specifications to add a new condition to extend the allowed completion time to restore one Essential Raw Cooling Water train to OPERABLE status from 72 hours to 7 days for planned maintenance, when the opposite unit is defueled or in Mode 6, following defueling under certain restrictions.

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

Response: No.

The proposed change adds new Condition A to Technical Specification (TS) 3.7.8, Essential Raw Cooling Water (ERCW) System for Sequoyah Nuclear Plant (SQN) Units 1 and 2. The proposed change will extend the allowed completion time to restore ERCW System train to OPERABLE status from 72 hours to 7 days for planned maintenance when the opposite unit is defueled or in mode 6 following defueled with refueling water cavity level ≥ [greater than or equal to] 23 ft. above top of reactor vessel flange and UHS [ultimate heat sink] Temperature is ≤ [less than or equal to 79 degrees F. This change does not result in any physical changes to plant safety-related structures, systems, or components (SSCs). The UHS and associated ERCW system function is to remove plant system heat loads during normal and accident conditions. As such, the UHS and ERCW system are not design basis accident initiators, but instead perform accident mitigation functions by serving as the heat sink for safety-related equipment to ensure the conditions and assumptions credited in the accident analyses are preserved. During operation under the proposed change with one ERCW train inoperable, the other ERCW train will continue to perform the design function of the ERCW system. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

Accordingly, as demonstrated by TVA design heat transfer and flow modeling calculations, operation with one ERCW System inoperable for 7 days for planned maintenance when the opposite unit is defueled or in mode 6 following defueled with refueling water cavity level ≥ 23 ft. above top of reactor vessel flange, the fuel cladding, Reactor Coolant System (RCS) pressure boundary, and containment integrity limits are not challenged during worst-case post-accident conditions. Accordingly, the conclusions of the accident analyses will remain as previously evaluated such that there will be no significant increase in the post-accident dose consequences.

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

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

Response: No.

The proposed change does not involve any physical changes to plant safety related SSCs or alter the modes of plant operation in a manner that is outside the bounds of the current UHS and ERCW system design heat transfer and flow modeling analyses. The proposed change to add new Condition A to TS 3.7.8, ERCW System, which would extend the allowed completion time to restore ERCW System train to OPERABLE status from 72 hours to 7 days for planned maintenance when the opposite unit is defueled or in mode 6 following defueled with refueling water cavity level ≥ 23 ft. above top of reactor vessel flange. Thus, although the specified ERCW system alignments result in reduced heat transfer flow capability, the plant's overall ability to reject heat to the UHS during normal operation, normal shutdown, and hypothetical worst-case accident conditions will not be significantly affected by this proposed change. Because the safety and design requirements continue to be met and the integrity of the RCS pressure boundary is not challenged, no new credible failure mechanisms, malfunctions, or accident initiators are created, and there will be no effect on the accident mitigating systems in a manner that would significantly degrade the plant's response to an accident.

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

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

Response: No.

The proposed change to add new Condition A to TS 3.7.8, ERCW System, which would extend the allowed completion time to restore ERCW System train to OPERABLE status from 72 hours to 7 days for planned maintenance when the opposite unit is defueled or in mode 6 following defueled with refueling water cavity level ≥ 23 ft. above top of reactor vessel flange. As demonstrated by TVA design basis heat transfer and flow modeling calculations, the design limits for fuel cladding, RCS pressure boundary, and containment integrity are not exceeded under both normal and post-accident conditions. As required, these calculations include evaluation of the worst-case combination of meteorology and operational parameters, and establish adequate margins to account for measurement and instrument uncertainties. While operating margins have been reduced by the proposed change in order to support necessary maintenance activities, the current limiting design basis accidents remain applicable and the analyses conclusions remain bounding such that the accident safety margins are maintained. Accordingly, the proposed change will not significantly degrade the margin of safety of any SSCs that rely on the UHS and ERCW system for heat removal to perform their safety related functions.

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.

NRC Branch Chief: Benjamin G. Beasley.

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

Date of amendment request: January 27, 2016. A publicly-available version is in ADAMS under Accession No. ML16033A470.

Description of amendment request: The amendment would revise the Technical Specifications to allow the use of Optimized ZIRLO^TM as an approved fuel rod cladding.

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

Response: No.

The proposed change would allow the use of Optimized ZIRLO^TM clad nuclear fuel in the reactor. The NRC approved topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, addresses Optimized ZIRLO^TM and demonstrates that Optimized ZIRLO^TM has essentially the same properties as currently licensed ZIRLO®. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO^TM fuel cladding will continue to meet the 10 CFR 50.46 acceptance criteria and, therefore, will not increase the consequences of an accident.

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

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

Response: No.

Use of Optimized ZIRLO^TM clad fuel will not result in changes in the operation or configuration of the facility. Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the material properties of Optimized ZIRLO^TM are similar to those of standard ZIRLO®. Therefore, Optimized ZIRLO^TM fuel rod cladding will perform similarly to those fabricated from standard ZIRLO®, thus precluding the possibility of the fuel cladding becoming an accident initiator and causing a new or different type of accident.

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

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

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 Optimized ZIRLO^TM are not significantly different from those of standard ZIRLO®. Optimized ZIRLO^TM is expected to perform similarly to standard ZIRLO® for all normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference in Optimized ZIRLO^TM material properties relative to standard ZIRLO® could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO^TM properties will demonstrate that the acceptance criteria of 10 CFR 50.46 have been satisfied.

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

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW., Washington, DC 20037.

NRC Branch Chief: Robert J. Pascarelli.

III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined 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.

A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated.

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

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the “Obtaining Information and Submitting Comments” section of this document.

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

Date of amendment request: June 30, 2015, as supplemented by letters dated August 11, 2015; September 24, 2015; October 8, 2015; December 7, 2015; February 10, 2016; and February 25, 2016.

Brief description of amendments: The amendments revised selected Technical Specification Completion Times to support repair activity associated with the Nuclear Service Water System, Train `A'.

Date of issuance: March 16, 2016.

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

Amendment Nos.: 282 and 261. A publicly-available version is in ADAMS under Accession No. ML15306A141; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-9 and NPF-18: Amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: August 20, 2015 (80 FR 50663). The supplemental letters dated August 11, 2015; September 24, 2015; October 8, 2015; December 7, 2015; February 10, 2016; and February 25, 2016, 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 amendments is contained in a Safety Evaluation dated March 16, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 24, 2015.

Brief description of amendment request: The amendment changed the VY Cyber Security Plan Implementation Schedule Milestone 8 full implementation date of June 30, 2016, to December 15, 2017. The amendment also revised the existing Renewed Facility Operating License Security Plan license condition.

Date of issuance: March 14, 2016.

Effective date: As of the date of issuance, and shall be implemented by June 30, 2015.

Amendment No.: 265. A publicly-available version is in ADAMS under Accession No. ML16014A169; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-28: The amendment revised the Facility Operating License.

Date of initial notice in Federal Register: September 8, 2015 (80 FR 53900).

The Commission's related evaluation of this amendment is contained in the Safety Evaluation dated March 14, 2016.

No significant hazards consideration comments received: No.

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

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: July 29, 2015, as supplemented by letter dated November 4, 2015.

Brief description of amendments: The amendments revised the emergency plan definition of annual training frequency to “once per calendar year not to exceed 18 months between training sessions.”

Date of issuance: March 17, 2016.

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

Amendment Nos.: 316/294; 221/155; and 121. A publicly-available version is in ADAMS under Accession No. ML15352A164; documents related to these amendments are listed in the safety evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-53, DPR-69, DPR-63, NPF-69, and DPR-18: The amendments revised the emergency plans.

Date of initial notice in Federal Register: December 8, 2015 (80 FR 76320).

The Commission's related evaluation of the amendments is contained in a safety evaluation dated March 17, 2016.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating Station (LGS), Unit 1, Montgomery County, Pennsylvania

Date of amendment request: November 19, 2015.

Brief description of amendment: The amendment revised the technical specifications (TSs) related to the safety limit minimum critical power ratios. The changes resulted from a cycle-specific analysis performed to support the operation of LGS, Unit 1, in the upcoming Cycle 17.

Date of issuance: March 15, 2016.

Effective date: As of the date of issuance and shall be implemented prior to startup from the spring 2016 refueling outage.

Amendment No.: 221. A publicly-available version is in ADAMS under Accession No. ML16041A021; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-39: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: January 5, 2016 (81 FR 275).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 15, 2016.

No significant hazards consideration comments received: No.

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 amendment request: September 4, 2014, as supplemented by letters dated January 29, February 6, April 28, July 6, September 4, October 1, and October 26, 2015, and January 15, 2016.

Brief description of amendments: The amendments changed the Technical Specifications (TSs) and Renewed Facility Operating Licenses (RFOLs) to allow plant operation from the currently licensed Maximum Extended Load Line Limit Analysis (MELLLA) domain to plant operation in the expanded MELLLA Plus (MELLLA+) domain under the previously approved extended power uprate conditions of 3,951 megawatts thermal rated core thermal power. The expanded MELLLA+ operating domain increases operating flexibility by allowing control of reactivity at maximum power by changing flow rather than by control rod insertion and withdrawal.

Date of issuance: March 21, 2016.

Effective date: As of the date of issuance and shall be implemented within 1 year of issuance.

Amendments Nos.: 305 and 309. A publicly-available version is in ADAMS under Accession No. s; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

RFOL Nos. DPR-44 and DPR-56: The amendments revised the RFOLs and TSs.

Date of initial notice in Federal Register: December 2, 2014 (79 FR 71454). The supplemental letters dated January 29, February 6, April 28, July 6, September 4, October 1, and October 26, 2015, and January 15, 2016, 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 amendments is contained in a Safety Evaluation dated March 21, 2016.

No significant hazards consideration comments received: Yes.

South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina

Date of amendment request: September 11, 2014, as supplemented by letters dated October 15, 2014, and December 18, 2014.

Description of amendment: The amendments revised the Updated Final Safety Analysis Report by clarifying how human diversity was applied during the design process for the Component Interface Module and Diverse Actuation System.

Date of issuance: July 17, 2015.

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

Amendment Nos.: 28. A publicly-available version is in ADAMS under Accession No. ML15176A703; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined License Nos. NPF-93 and NPF-94: Amendments revised the Facility Combined Licenses.

Date of initial notice in Federal Register: December 9, 2014 (79 FR 73111). The supplemental letters dated October 15, 2014, and December 18, 2014, 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 amendments is contained in a Safety Evaluation dated July 17, 2015.

No significant hazards consideration comments received: No.

South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina

Date of amendment request: February 10, 2015.

Brief description of amendment: The amendments revised the VCSNS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) by revising the references to human factors-related plans. The UFSAR-referenced plans are the Human Factors Engineering Design Verification plan, Task Support Verification plan, and the Integrated System Validation plan. The UFSAR references to those plans required an update to the latest version of those plans due to changes within the plans. The amendments involved changes to the approved VCSNS Units 2 and 3 UFSAR Tier 2* information, as defined in 10 CFR part 52, appendix D, section II.F.

Date of issuance: September 23, 2015.

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

Amendment Nos.: 33. A publicly-available version is in ADAMS under Accession No. ML15189A363; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined License Nos. NPF-93 and NPF-94: Amendments revised the Facility Combined Licenses.

Date of initial notice in Federal Register: March 31, 2015 (80 FR 17094). The supplemental letter dated March 24, 2015, 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 amendments is contained in a Safety Evaluation dated September 23, 2015.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina

Date of amendment request: August 24, 2015.

Brief description of amendment: The amendments authorized changes to the VCSNS Units 2 and 3 Updated Final Safety Analysis Report Tier 2 and Tier 2* information to revise the seismic Category I and II structures containing mechanical couplers welded to structural steel utilizing combined partial joint penetration weld with fillet weld reinforcement with fillet welds satisfying the minimum size requirements for C2/C3J couplers to demonstrate the capacity required by code is established by appropriate analyses and testing.

Date of issuance: November 12, 2015.

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

Amendment Nos.: 36. A publicly-available version is in ADAMS under Accession No. ML15301A100; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined License Nos. NPF-93 and NPF-94: Amendments revised the Facility Combined Licenses.

Date of initial notice in Federal Register: September 3, 2015 (80 FR 53336). The supplemental letters dated September 23, 2015, and October 1, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 12, 2015.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina

Date of amendment request: October 22, 2015.

Brief description of amendment: The amendments authorized changes to the VCSNS Combined Licenses (COLs). Specifically, the changes were to VCSNS Units 2 and 3 COLs, Appendix A, Technical Specifications, Section 5.0, “Administrative Controls,” by revising the title “Shift Supervisor” to “Shift Manager.”

Date of issuance: February 29, 2016.

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

Amendment Nos.: 42. A publicly-available version is in ADAMS under Accession No. ML16042A476; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined License Nos. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: November 24, 2015 (80 FR 73242).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 29, 2016.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: May 12, 2015, as supplemented by letters dated September 21, 2015; November 25, 2015; and January 28, 2016.

Brief description of amendments: The amendments revised and added Surveillance Requirements to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances that permit performance of the verification. The changes are consistent with TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”

Date of issuance: March 21, 2016.

Effective date: As of the date of issuance and shall be implemented within 120 days of issuance.

Amendment Nos.: 178 (Unit 1) and 159 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16063A475, documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: June 23, 2015 (80 FR 35984). The supplemental letters dated September 21, 2015; November 25, 2015; and January 28, 2016, 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 amendments is contained in a Safety Evaluation dated March 21, 2016.

No significant hazards consideration comments received: No.

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, Unit Nos. 1 and 2, Appling County, Georgia

Date of amendment request: April 2, 2015, as supplemented by letters dated November 12, 2015, and February 9, 2016.

Brief description of amendments: The amendments revised the technical specifications (TSs) as necessary to relocate the pressure and temperature (P-T or P/T) limit curves and associated references to a pressure and temperature limits report (PTLR). Specifically, the request modified Section 1.0, “Definitions”; Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9, “RCS Pressure and Temperature (P/T) Limits”; and Section 5.0, “Administrative Controls,” of the TSs for both units to delete reference to the P-T curves and to include reference to the unit-specific PTLRs. The amendments also implemented new P-T limits for both units.

Date of issuance: March 23, 2016.

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

Amendment Nos.: 277 and 221. A publicly-available version is in ADAMS under Accession No. ML16062A099; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: July 7, 2015 (80 FR 38760). The supplemental letters dated November 12, 2015, and February 9, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 23, 2016.

No significant hazards consideration comments received: No.

Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of amendment request: March 19, 2015, as supplemented by letters dated October 15, 2015; October 16, 2015; and January 8, 2016. A publicly-available version is in ADAMS under Accession Nos. ML15091A657, ML15296A048, ML15296A057, and ML16011A103, respectively.

Brief description of amendments: The amendments revised the Emergency Plan for the Susquehanna Steam Electric Station (SSES) to adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action Level scheme described in NEI 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors” (ADAMS Accession No. ML12326A805), which was endorsed by the NRC as documented in NRC letter dated March 28, 2013 (ADAMS Accession No. ML12346A463). This request was submitted by PPL Susquehanna, LLC; however, on June 1, 2015 (ADAMS Accession No. ML15054A066), the NRC staff issued an amendment changing the name on the SESS license from PPL Susquehanna, LLC to Susquehanna Nuclear, LLC. This amendment was issued subsequent to an order issued on April 10, 2015 (ADAMS Accession No. ML15058A073), to SSES, approving an indirect license transfer of the SESS license to Talen Energy Corporation.

Date of issuance: March 28, 2016.

Effective date: As of the date of issuance and shall be implemented on or before December 31, 2016.

Amendment Nos.: 265 (Unit 1) and 246 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16062A216; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: July 7, 2015 (80 FR 38762). The supplemental letters dated October 15, 2015; October 16, 2015; and January 8, 2016, provided additional information that clarified the application and expanded the scope of the application as originally noticed, and changed the staff's original proposed no significant hazards consideration determination as published in the Federal Register. As such, the NRC staff published a subsequent notice in the Federal Register on February 2, 2016 (81 FR 5500).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 28, 2016.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 1st day of April 2016.

For the Nuclear Regulatory Commission.

Anne T. Boland,

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

[FR Doc. 2016-08323 Filed 4-11-16; 8:45 am]

BILLING CODE 7590-01-P