81 Fed. Reg. 28891 (May. 10, 2016)

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

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Federal RegisterMay 10, 2016

81 Fed. Reg. 28891 (May. 10, 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 April 12 to April 25, 2016. The last biweekly notice was published on April 26, 2016 (81 FR 24659).

DATES:

Comments must be filed by June 9, 2016. A request for a hearing must be filed by July 11, 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-2016-0093. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
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:

Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov.

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2016-0093 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-2016-0093.
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.
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-2016-0093, 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 will post all comment submissions at http://www.regulations.gov as well as enter 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 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)-(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 July 11, 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 July 11, 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 ten 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.

Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power Station, Unit No. 1 (MPS1), New London County, Connecticut

Date of amendment request: March 28, 2014. A publicly-available version is in the ADAMS under Accession No. ML14093A028.

Description of amendment request: The amendment would make changes to the MPS1 Permanently Defueled Technical Specifications (PDTSs) by deleting the Table of Contents section and making administrative changes to the PDTSs.

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

1) Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes are administrative in nature. The proposed changes remove the PDTS Table of Contents section and make two other administrative changes to the PDTSs. Furthermore, MPS1 has permanently ceased operation and is being maintained in a defueled condition. Therefore, the only credible design basis accident is a fuel handling accident. The administrative changes proposed herein are not initiators of any fuel handling accident previously evaluated, and, consequently, the probability and consequences of a fuel handling accident previously evaluated is not significantly increased.

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

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

Response: No.

The proposed changes are administrative in nature, therefore no new or different accidents result from the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The changes do not alter assumptions made in the safety analysis.

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

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

Response: No.

The proposed administrative changes do not involve a change in the method of plant operation, do not affect any accident analyses, and do not relax any safety system settings.

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

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

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

NRC Branch Chief: Bruce A. Watson.

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: February 18, 2016. A publicly-available version is in ADAMS under Accession No. ML16076A413.

Description of amendment request: The amendment would allow a one-time extension to the 10-year frequency of the McGuire Nuclear Station, Units 1 and 2, containment leakage rate tests. The change would extend the period from 10 years to 10.5 years for each unit.

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

Response: No.

The proposed amendment to the Technical Specifications (TS) involves the extension of the McGuire Nuclear Station (MNS) Type A containment integrated leak rate test interval to 10.5 years. The current Type A test interval of 120 months (10 years) would be extended on a one-time basis to no longer than 10.5 years from the last Type A test. This extension is bounded by the 15 month extension, permissible only for non-routine emergent conditions, allowed in accordance with NEI [Nuclear Energy Institute] 94-01 revision 0. The proposed extension also does not change the test method or procedure. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. The containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from 10 years to 10.5 years, measured, as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.023 person-rem/year. EPRI [Electric Power Research Institute] Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of ≤ 1.0 person-rem per year, or ≤ 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT [integrated leak rate test] intervals. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

As documented in NUREG-1493, Performance-Based Containment Leak-Test Program, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The MNS Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and; (2) time based as previously discussed. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME Section XI, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.

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

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

Response: No.

The proposed amendment to the TS involves the extension of the MNS Type A containment integrated leak rate test interval from 10 years to 10.5 years. The current Type A test interval of 120 months (10 years) would be extended on a one-time basis to 10.5 years from the last Type A test. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

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

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

Response: No.

The proposed amendment to TS 5.5.2 involves the extension of the MNS Type A containment integrated leak rate test interval to 10.5 years. The current Type A test interval of 120 months (10 years) would be extended on a one-time basis to no longer than 10.5 years from the last Type A test. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type A containment leak rate tests for MNS. The proposed surveillance interval extension is bounded by the 15-year ILRT interval currently authorized within NEI 94-01, Revisions 2-A and 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the approval of this proposed change, since these are not affected by changes to the Type A test intervals.

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

Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: January 25, 2016, as supplemented by letter dated March 31, 2016. A publicly-available version is in ADAMS under Accession Nos. ML16025A182 and ML16076A077.

Description of amendment request: The proposed amendment would revise the technical specifications (TSs) to allow a permanent extension of the Type “A” integrated leak rate testing and Type “C” leak rate testing frequencies. This request also proposes to delete information in TS 5.5.13 regarding a completed requirement to perform Type “C” testing in 2008.

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

Response: No.

The proposed activity involves the extension of the Clinton Power Station (CPS), Unit 1, Type A containment test interval to 15 years, and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

The change in dose risk for changing the Type A Integrated Leak Rate Test (ILRT) interval from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI [Electric Power Research Institute] guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr. The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

In addition, as documented in NUREG-1493, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The CPS, Unit 1 Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and, (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Section XI, and Technical Specifications (TS) requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for CPS. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how CPS is operated.

Therefore, the proposed change does not result in 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 amendment to the TS 5.5.13, “Primary Containment Leakage Rate Testing Program,” involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident.

The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.

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 proposed amendment to TS 5.5.13 involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leaktightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves the extension of the interval between Type A containment leak rate tests and Type C tests for CPS, Unit 1. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI [Nuclear Energy Institute] 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for CPS, Unit 1. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how CPS is operated and maintained. Thus, there is no reduction in any margin of safety.

Therefore, the proposed changes do 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.

NRC Acting Branch Chief: Justin C. Poole.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: February 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16054A359.

Description of amendment request: The amendment would revise the Technical Specifications to incorporate previously NRC-approved Industry/Technical Specification Task Force 439 (TSTF-439), Revision 2, “Eliminate Second Completion Times Limiting Time From Discovery of Failure To Meet an LCO.”

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

Response: No.

The proposed change eliminates certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the revised Completion Time are no different than the consequences of the same accident during the existing Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. The proposed change does not alter or prevent the ability of SSCs [systems, structures, and components] from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change 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. The proposed change does not alter any assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of anew 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 delete the second Completion Time does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside of the design 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 Road, Warrenville, IL 60555.

NRC Branch Chief: Travis L. Tate.

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

Date of amendment request: March 18, 2016. A publicly-available version is in ADAMS under Accession No. ML16078A065.

Description of amendment request: The amendment would revise the Technical Specifications (TS) concerning a change to the method of calculating core reactivity for the purpose of performing the Reactivity Anomalies surveillance at NMPNS, Units 1 and 2.

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

Response: No.

The proposed TS changes do not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The amendment would only change how the Reactivity Anomalies surveillance is performed. Verifying that the core reactivity is consistent with predicted values ensures that accident and transient safety analyses remain valid. This amendment changes the TS requirements such that, rather than performing the surveillance by comparing predicted to actual control rod density, the surveillance is performed by a direct comparison of k_eff.

Therefore, since the Reactivity Anomalies surveillance will continue to be performed by a viable method, the proposed amendment does not involve a significant increase in the probability or consequence of a previously evaluated accident.

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

Response: No.

This TS amendment request does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. The proposed changes to the Reactivity Anomalies surveillance will only provide a new, more efficient method of detecting an unexpected change in core reactivity.

Since all systems continue to be operated within their design bases, no new failure modes are introduced and the possibility of a new or different kind of accident is not created.

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

Response: No.

This proposed TS amendment proposes to change the method for performing the Reactivity Anomalies surveillance from a comparison of predicted to actual control rod density to a comparison of predicted to monitored k_eff. The direct comparison of k_eff provides a technically superior method of calculating any differences in the expected core reactivity. The Reactivity Anomalies surveillance will continue to be performed at the same frequency as is currently required by the TS, only the method of performing the surveillance will be changed. Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified.

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

NRC Branch Chief: Travis L. Tate.

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

Date of amendment request: February 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16055A149.

Description of amendment request: The amendment would (1) revise Technical Specification (TS) 4.2.1, “Reactor Core, Fuel Assemblies,” to add Optimized ZIRLO^TM, as an approved fuel rod cladding material, (2) revise TS 5.6.5.b to add the Westinghouse topical reports for Optimized ZIRLO^TM and ZIRLO®, and (3) revise TS 5.6.5.b with a non-technical change to the Reference 11 title (replace a semicolon with a period).

EGC [Exelon Generation Company] has evaluated the proposed changes for Braidwood and Byron, using the criteria in 10 CFR 50.92, and has determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

Criteria

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 reactors. The NRC approved topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO^TM prepared by Westinghouse Electric Company LLC (Westinghouse), 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. With the approved exemption, use of Optimized ZIRLO^TM fuel cladding will continue to meet all 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 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. Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, 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 is 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.

Based on the above, EGC concludes that the proposed amendment to allow the use of Optimized ZIRLO^TM fuel cladding material does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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

NRC Acting Branch Chief: Justin C. Poole.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: March 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16075A411.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.6.2.2, “Suppression Pool Water Level,” as well as TS surveillance requirements 3.6.2.4.1 and 3.6.2.4.4 associated with TS 3.6.2.4, “Suppression Pool Makeup System (SPMU),” to allow installation of the reactor well to steam dryer storage pool gate in the upper containment pool (UCP) in MODES 1, 2, and 3. The proposed amendment would also create new special operations TS 3.10.9, “Suppression Pool Makeup—MODE 3 Upper Containment Pool Drain-Down,” to allow draining of the reactor well portion of the UCP in MODE 3.

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

Response: No.

The changes proposed in the license amendment request specify different water level requirements in the upper containment pool and suppression pool to permit gate installation in MODES 1, 2, and 3, and drain-down of the reactor well in MODE 3. The probability of an accident previously evaluated is unrelated to the water level in these pools, since they are mitigating systems. The operation or failure of a mitigating system does not contribute to the occurrence of an accident. No active or passive failure mechanisms that could lead to an accident are affected by these proposed changes.

Suppression pool water levels are increased during upper pool gate installation in MODES 1, 2, and 3 and during reactor well drain-down in MODE 3, with a potential for an increased probability of drywell flooding during an inadvertent dump of the upper containment pool. An inadvertent dump of the upper pool during any period of operation with a pressurized vessel does not represent, in and of itself, any significant hazard to the public, the plant operating personnel, or any plant equipment. The piping components which would be affected in this event have been analyzed for the flooding effect, and it has been determined that this event could not initiate a loss of coolant accident (LOCA).

The changes have no impact on the ability of any of the emergency core cooling systems (ECCS) to function adequately, since adequate net positive suction head (NPSH) is maintained. The increase in suppression pool water level to compensate for the reduction in UCP volume will provide reasonable assurance that the minimum post-accident vent coverage is adequate to assure the pressure suppression function of the suppression pool is accomplished. The suppression pool water level will be raised above the current high water level for the proposed reactor well drain-down activity only after the reactor pressure has been reduced sufficiently to assure that the hydrodynamic loads from a loss of coolant accident will not exceed the design values. The reduced reactor pressure will also ensure that the loads due to main steam safety relief valve actuation with an elevated pool level are within the design loads.

Relative to dose rates on the refuel floor, the resultant dose rates from the reactor in MODES 3 and 4 are the same regardless of a drain-down of the upper pool reactor well. Relative to a low pressure LOCA in MODE 3, the reduced post-LOCA containment pressure and the decay time to reach MODE 3 conditions ensures that post-accident dose consequences are bounded by the design-basis accident LOCA.

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

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

Response: No.

The proposed changes specify different water level requirements in the upper containment pool and suppression pool to permit gate installation in MODES 1, 2, and 3, and drain-down of the reactor well in MODE 3. These changes do not affect or alter the ability of the suppression pool makeup (SPMU) system to perform its design function. The proposed change in the pool water levels will maintain the design function of mitigating the pressure and temperature increase generated by a LOCA, and will maintain the required drywell vent coverage during post-accident ECCS draw down.

The altered water levels in the pools do not create a different type of accident than presently evaluated. With the reduced pressure in the reactor coolant system, the GOTHIC computer program simulations demonstrate that the accident responses at defined conditions with the reactor well drained in MODE 3 are bounded by the current design basis accidents.

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

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

Response: No.

The proposed changes to the UCP and the suppression pool water levels do not introduce any new setpoints at which protective or mitigating actions are initiated. Current instrument setpoints remain unaltered by this change. Although the water levels are adjusted for the UCP gate installation and the reactor well drain-down activity, the design and functioning of the containment pressure suppression system remains unchanged. The proposed total water volume is sufficient to provide high confidence that the pressure suppression and containment systems will be capable of mitigating large and small break accidents. All analyzed accident results remain within the design values for the structures and equipment.

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

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

NRC Branch Chief: David J. Wrona.

Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, California

Date of amendment request: March 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16084A588.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.4.12, “Low Temperature Overpressure Protection (LTOP) System,” to reflect the mass input transient analysis that assumes an emergency core cooling system (ECCS) centrifugal charging pump (CCP) and the normal charging pump (NCP) capable of simultaneously injecting into the reactor coolant system (RCS) during TS 3.4.12 applicability.

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 TS 3.4.12 to allow an ECCS CCP and the NCP aligned to LTOP orifice to be capable of injecting into the RCS during low RCS pressures and temperatures. The LCO [Limiting Condition for Operation] provides RCS overpressure protection by having a minimum coolant input capability and have adequate pressure relief capability. Analyses have demonstrated that one power operated relief valve (PORV) or an RCS vent of at least 2.07 square inches is capable of limiting the RCS pressure excursions below the 10 CFR 50, Appendix G limits for the design basis LTOP limits.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed change does not adversely affect the ability of structures, systems, and components to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed change does not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure.

The NRC has previously evaluated a similar LAR [license amendment request] related to Wolf Creek Generating Station. In Amendment No. 207, the NRC concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated [ADAMS Accession No. ML13282A534].

In 2007, PG&E replaced the Unit 1 non-safety-related PDP [positive displacement pump] with a non-safety-related CCP, called the NCP, in order to alleviate operational issues associated with the PDP. In 2008, PG&E performed the replacement on Unit 2. PG&E also designed, tested, and installed an FCO [flow choking orifice] called the LTOP orifice to be used during LTOP operation to ensure that the total maximum mass injection capability with the NCP remained bounded by the LTOP mass injection analysis. These changes were implemented under 10 CFR 50.59. However, no physical changes are being made to the plant as a result of the proposed license amendment.

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 accident from any accident previously evaluated?

Response: No.

The proposed change revises TS 3.4.12 to allow an ECCS CCP and the NCP aligned to LTOP orifice to be capable of simultaneously injecting into the RCS during low RCS pressures and temperatures. The LCO provides RCS overpressure protection by having a minimum coolant input capability and have adequate pressure relief capability. Analyses have demonstrated that one PORV or an RCS vent of at least 2.07 square inches is capable of limiting the RCS pressure excursions below the 10 CFR 50, Appendix G limits for the design basis LTOP limits.

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

Therefore, the proposed change does not create the possibility of a new or different 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 does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside the design basis.

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

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, CA 94120.

NRC Branch Chief: Robert J. Pascarelli.

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: March 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16067A145.

Description of amendment request: The proposed changes, if approved, would amend Combined License (COL) No. NPF-93 and NPR-94 for the VCSNS. The requested amendment proposed changes would depart from the approved AP1000 Design Control Document (DCD) “Tier 2” and “Tier 2*” information as currently incorporated into the VCSNS Updated Final Safety Analysis Report (UFSAR). The changes relate to updating the UFSAR text and tables; and information incorporated by reference related to Westinghouse Electric Company's Reports WCAP-16096, “Software Program Manual for Common Q^TM Systems,” (also known as the Common Q SPM) Revision 4, WCAP-16097, “Common Qualified Platform Topical Report,” (also known as the Common Q Topical Report) Revision 3, and WCAP-15927, “Design Process for AP1000 Common Q Safety Systems,” Revision 4; and associated documents and references such as a reference to the NRC's Regulatory Guide 1.152, “Criteria for Use of Computers in Safety Systems of Nuclear Power Plants” (Revision 3, July 2011), and its associated exceptions. The proposed changes also include removal of Tier 2* WCAP-17201-P, “AC160 High Speed Link Communication Compliance to DI&C-ISG-04 Staff Positions 9, 12, 13 and 15 Technical Report,” as a UFSAR incorporated by reference document.

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 Updated Final Safety Analysis Report. 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 (IEEE) 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 Protection and Safety Monitoring System (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 (SSC) accident initiator or initiating sequence of events. Thus, the probabilities of the accidents previously evaluated in the UFSAR are not affected.

Therefore, the proposed activity 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 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 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: Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.

NRC Acting Branch Chief: John McKirgan.

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: March 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16075A264.

Description of amendment request: The proposed change would amend the Combined License (COL) No. NPF-93 and NPF-94 for the VCSNS. The requested amendment proposes to depart from approved AP1000 Design Control Document (DCD) Tier 2 information (text, tables, and figures) and involved Tier 2* information (as incorporated into the Updated Final Safety Analysis Report as plant specific DCD information), and also involves a change to the plant-specific Technical Specifications. Specifically, the amendment request proposes changes to the plant-specific AP1000 fuel system design, nuclear design, thermal hydraulic design, and accident analyses as described in the licensing basis documents. These proposed changes are consistent with those generically approved in WCAP-17524-P-A, Revision 1, “AP1000 Core Reference Report.”

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

Response: No.

The proposed changes will revise the licensing basis documents related to the fuel system design, nuclear design, thermal hydraulic design, and accident analyses.

The UFSAR [Updated Final Safety Analysis Report] Chapter 15 accident analyses describe the analyses of various design basis transients and accidents to demonstrate compliance of the AP1000 design with the acceptance criteria for these events. The acceptance criteria for the various events are based on meeting the relevant regulations, general design criteria, the Standard Review Plan, and are a function of the anticipated frequency of occurrence of the event and potential radiological consequences to the public. As such, each design-basis event is categorized accordingly based on these considerations. As discussed in Section 5.3 of WCAP-17524-P-A Revision 1, the revised accident analyses maintain their plant conditions, and thus their frequency designation and consequence level as previously evaluated. As confirmed in the Safety Evaluation Report (SER), the revised analyses meet the applicable guidelines in the Standard Review Plan.

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

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

Response: No.

The proposed changes will revise the licensing basis documents related to the fuel system design, nuclear design, thermal hydraulic design, and accident analyses.

The proposed changes would not introduce a new failure mode, fault, or sequence of events that could result in a radioactive material release. The proposed changes do not alter the design, configuration, or method of operation of the plant beyond standard functional capabilities of the equipment.

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

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

Response: No.

The proposed changes will revise the licensing basis documents related to the fuel system design, nuclear design, thermal hydraulic design, and accident analyses.

Safety margins are applied at many levels to the design and licensing basis functions and to the controlling values of parameters to account for various uncertainties and to avoid exceeding regulatory or licensing limits. UFSAR Subsection 4.1.1 presents the Principle Design Requirements imposed on the fuel and control rod mechanism design to ensure that the performance and safety criteria described in UFSAR Chapter 4 and Chapter 15 are met. The revised fuel system design, nuclear design, thermal hydraulic design, and accident analyses maintain the same Principle Design Requirements, and further, satisfy the applicable regulations, general design criteria, and Standard Review Plan. The effects of the changes do not result in a significant reduction in margin for any safety function, and were evaluated in the Safety Evaluation Report for WCAP-17524-P-A Revision 1 and found to be acceptable.

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

Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.

NRC Acting Branch Chief: John McKirgan.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee

Date of amendment request: February 23, 2016. A publicly-available version is in ADAMS under Accession No. ML16054A585.

Description of amendment request: The amendment would revise the WBN Dual Unit Fire Protection Report and would revise the associated License Condition regarding the WBN fire protection program. Specifically, the amendment requests approval of a deviation from the physical separation requirements of 10 CFR part 50, appendix R, section III.G.2.d.

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

Response: No.

A fire hazards analysis was performed for the areas under the scope of this amendment. This fire hazards analysis demonstrates that one train of safe shutdown equipment will remain functional in the event of an Appendix R fire, even though a radiant energy shield will not be provided for two raceway containing safe shutdown circuits.

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

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

Response: No.

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.

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

Attorney for licensee: Sherry A. Quirk, Executive Vice President and General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, Knoxville, TN 37902.

NRC Branch Chief: Benjamin G. Beasley.

III. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: March 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16064A488.

Brief description of amendment request: The amendment would revise the Cyber Security Plan implementation schedule for Milestone 8 and would revise the associated license condition in the Facility Operating License.

Date of publication of individual notice in Federal Register: April 19, 2016 (81 FR 23011).

Expiration date of individual notice: May 19, 2016 (public comments); June 20, 2016 (hearing requests).

IV. 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.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: September 24, 2015.

Brief description of amendment: The amendment revises Surveillance Requirements (SRs) to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances which permit performance of the verification. The changes address the concerns discussed in NRC Generic Letter (GL) 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” as described in NRC-approved Technical Specifications Task Force (TSTF)-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”

Date of issuance: April 20, 2016.

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

Amendment No.: 204. A publicly-available version is in ADAMS under Accession. No. ML16069A006; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-43: This amendment revises the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: January 5, 2016 (81 FR 260).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 20, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 30, 2015, as supplemented by letter dated February 19, 2016.

Brief description of amendments: The amendments approved adoption of an emergency action level scheme based on Nuclear Energy Institute (NEI) 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” for the Catawba Nuclear Station, Units 1 and 2.

Date of issuance: April 18, 2016.

Effective date: As of the date of issuance and shall be implemented by March 10, 2017.

Amendment Nos.: 279 for Unit 1 and 275 for Unit 2. A publicly-available version is in ADAMS under Accession No. ML16082A038; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: The amendments revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: June 23, 2015 (80 FR 35980). The supplemental letter dated February 19, 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 April 18, 2016.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-369, 50-370, 50-413, and 50-414, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina and Catawba Nuclear Station, Units 1 and 2, York County, SC

Date of amendment request: June 23, 2015.

Brief description of amendments: The amendments remove superseded TS requirements.

Date of issuance: April 8, 2016.

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

Amendment Nos.: 283, 262, 278, and 274. A publicly-available version is in ADAMS under Accession No. ML16060A229; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-9, NPF-17, NPF-35, and NPF-52: Amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: August 4, 2015 (80 FR 46347).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 8, 2016.

No significant hazards consideration comments received: No.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of amendment request: April 30, 2015, as supplemented by letters dated November 19, 2015, and January 28, 2016.

Brief description of amendment: The amendment adopted the NRC-endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, “Methodology for the Development of Emergency Action Levels for Non-Passive Reactors.”

Date of issuance: April 13, 2016.

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

Amendment No.: 149. A publicly-available version is in ADAMS under Accession No. ML16057A838; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment.

Facility Operating License No. NPF-63: The amendment revised the Emergency Action Level Technical Bases document.

Date of initial notice in Federal Register: July 21, 2015 (80 FR 43128). The supplemental letters dated November 19, 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 amendment is contained in an SE dated April 13, 2016.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-286, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3, Westchester County, New York

Date of amendment request: June 16, 2015.

Brief description of amendments: The amendments revised the Cyber Security Plan Milestone 8 full implementation date by extending the full implementation date from June 30, 2016, to December 31, 2017.

Date of issuance: April 12, 2016.

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

Amendment Nos.: 59 (Unit No. 1), 284 (Unit No. 2), and 260 (Unit No. 3). A publicly-available version is in ADAMS under Accession No. ML16064A215; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Provisional Operating License No. DPR-5 and Facility Operating License Nos. DPR-26 and DPR-64: The amendments revised the Provisional Operating License for Unit No. 1 and the Facility Operating Licenses for Unit Nos. 2 and 3.

Date of initial notice in Federal Register: August 4, 2015 (80 FR 46348).

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

Date of amendment request: November 5, 2015.

Brief description of amendments: The amendments revise the Surveillance Requirement (SR) frequencies for SRs 3.4.6.4, 3.4.7.4, 3.4.8.3, 3.5.2.10, 3.6.6.9, 3.9.4.2, and 3.9.5.4. The changes to the SR frequencies relocate the frequencies to the Surveillance Frequency Control Program.

Date of issuance: April 11, 2016.

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

Amendment Nos.: 317 and 295. A publicly-available version is in ADAMS under Accession No. ML16060A401; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: January 5, 2016 (81 FR 261).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 11, 2016.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: March 23, 2015, as supplemented by letters dated January 8, 2016, and March 21, 2016.

Brief description of amendment: The amendment revised the technical specifications (TS) and relocated the secondary containment bypass leakage paths table from the TS to the Technical Requirements Manual.

Date of issuance: April 19, 2016.

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

Amendment No.: 156. A publicly-available version is in ADAMS under Accession No. ML16088A053; documents related to this amendment is listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-69: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: September 29, 2015 (80 FR 58517). The supplemental letters dated January 8, 2016, and March 21, 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 April 19, 2016.

No significant hazards consideration comments received: No.

Florida Power & Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2 (PSL-2), St. Lucie County, Florida

Date of amendment request: December 30, 2014, as supplemented by letters dated March 23, June 2, June 18, July 30, October 2, November 3, 2015; and December 8, 2015.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) to allow the use of AREVA fuel and AREVA M5® material as an approved fuel rod cladding at PSL-2.

Date of issuance: April 19, 2016.

Effective date: As of the date of issuance and shall be implemented upon the start of the PSL-2 Cycle 23 spring 2017 refueling outage to support the AREVA fuel transition project plan.

Amendment No.: 182. A publicly-available version is in ADAMS under Accession No. ML16063A121; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-16: Amendment revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: June 9, 2015 (80 FR 32620). The supplements dated June 2, June 18, July 30, October 2, November 3, and December 8, 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 April 19, 2016.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendments: June 26, 2013, as supplemented by letters dated September 29, October 27, October 29, November 26, and December 31, 2014; February 25 (two letters), May 7, October 15, and December 31, 2015; and January 28, 2016.

Brief description of amendments: The amendments permit the PG&E (the licensee) to adopt a new fire protection licensing basis based on National Fire Protection Association (NFPA) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition),” at Diablo Canyon Power Plant, Units 1 and 2, that complies with the requirements of 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide 1.205, “Risk Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,” December 2009.

Date of issuance: April 14, 2016.

Effective date: As of its date of issuance and shall be implemented as described in the transition license conditions.

Amendment Nos.: Unit 1—225; Unit 2—227. A publicly-available version is in ADAMS under Accession No. ML16035A441; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: December 26, 2013 (78 FR 78408). The supplemental letters dated October 3, 2013; September 29, October 27, October 29, November 26, and December 31, 2014; February 25 (two letters), May 7, October 15, and December 31, 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 April 14, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 1, 2015.

Brief description of amendment: The amendment authorized changes to the VEGP Units 3 and 4 plant specific emergency planning inspections, tests, analyses, and acceptance criteria (ITAAC) in Appendix C of VEGP Units 3 and 4 Combined Operating Licenses (COLs). The changes authorize the removal of the copy of Updated Final Safety Analysis Report Table 7.5-1, “Post-Accident Monitoring System” from ITAAC in Appendix C of the VEGP Units 3 and 4 COLs.

Date of issuance: March 30, 2016.

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

Amendment No.: 47. A publicly-available version is in ADAMS under Accession No. ML16061A220; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: October 27, 2015 (80 FR 65807).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 30, 2015.

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: January 13, 2015, as supplemented by letters dated June 16 and November 24, 2015.

Brief description of amendments: The amendments adopt Technical Specification Task Force change number 523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation,” for the Hatch Nuclear Plant, Unit Nos 1 and 2, technical specifications. The change revised or added surveillance requirements to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances which permit performance of the verification.

Date of issuance: April 14, 2016.

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

Amendment Nos.: 278 and 222. A publicly-available version is in ADAMS under Accession No. ML16090A174; 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 Technical Specifications.

Date of initial notice in Federal Register: March 17, 2015 (80 FR 13911). The supplemental letters dated June 16 and November 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 amendment is contained in a Safety Evaluation dated April 14, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 23, 2015.

Brief description of amendment: The amendment revised the diesel generator (DG) full load rejection test and endurance and margin test specified by Technical Specification (TS) 3.8.1, “AC [Alternating Current] Sources—Operating,” Surveillance Requirements (SR) 3.8.1.10 and 3.8.1.14, respectively. The change adds a new Note to SR 3.8.1.10 and SR 3.8.1.14, consistent with Technical Specification Task Force (TSTF) traveler TSTF-276-A, Revision 2, “Revise DG full load rejection test.” The Note allows the full load rejection test and endurance and margin test to be performed at the specified power factor with clarifications addressing situations when the power factor cannot be achieved.

Date of issuance: April 15, 2016.

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

Amendment No.: 215. A publicly-available version is in ADAMS under Accession No. ML16081A194; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

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

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

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 2nd day of May 2016.

For the Nuclear Regulatory Commission.

Anne T. Boland,

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

[FR Doc. 2016-10949 Filed 5-9-16; 8:45 am]

BILLING CODE 7590-01-P