81 Fed. Reg. 92863 (Dec. 20, 2016)

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

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Federal RegisterDec 20, 2016

81 Fed. Reg. 92863 (Dec. 20, 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 November 22, 2016, to December 5, 2016. The last biweekly notice was published on December 6, 2016.

DATES:

Comments must be filed by January 19, 2017. A request for a hearing must be filed by February 21, 2017.

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-2016-0256. 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.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2016-0256, facility name, unit number(s), plant docket number, application date, and subject 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-2016-0256.
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 this document.
NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2016-0256, facility name, unit number(s), plant docket number, 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, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

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

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if 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. If the Commission takes 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. If the Commission makes 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 persons (petitioner) whose interest may be affected by this action may file a request for a hearing and a petition to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested persons 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 petition 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 petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition 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 petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the 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 petitioner's interest. The petition must also set forth the specific contentions which the 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 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 petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion to support its position on the issue. 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 proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A 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 consistent with the NRC's regulations, policies, and procedures.

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

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of any amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1).

The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by February 21, 2017. 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 set forth in this section, except that under 10 CFR 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. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.

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 (hereinafter “petition”), 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, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request: (1) A digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition (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 available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/adjudicatory-sub.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 Electronic Filing Help Desk will not be able to offer assistance in using unlisted software.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a petition. Submissions should be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.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 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 Electronic Filing 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 Electronic Filing Help Desk is available between 9 a.m. and 7 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 stating why there is good cause for not filing electronically and 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, 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 petition 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.

The Commission will issue a notice or order granting or denying a hearing request or intervention petition, designating the issues for any hearing that will be held and designating the Presiding Officer. A notice granting a hearing will be published in the Federal Register and served on the parties to the hearing.

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

Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of amendment request: September 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16287A421.

Description of amendment request: The amendment would revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) values contained in the Technical Specifications (TSs) for two recirculation loop operation and for single loop recirculation operation.

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

Response: No.

The proposed SLMCPR values have been determined using NRC-approved methods discussed in AREVA Topical Report ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, June 2011, as augmented by the associated TS Appendix B Additional Condition related to channel bow model uncertainty. Establishing a two recirculation loop SLMCPR value of ≥1.07 and a single recirculation loop SLMCPR value of ≥1.09 ensures that the acceptance criteria continues to be met (i.e., at least 99.9 percent of all fuel rods in the core do not experience transition boiling).

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The proposed license amendments do not involve any plant modifications or operational changes that could affect system reliability or performance, or that could affect the probability of operator error. As such, the proposed changes do not affect any postulated accident precursors. Since no individual precursors of an accident are affected, the proposed license amendments do not involve a significant increase in the probability of a previously analyzed event.

The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. The basis for the SLMCPR calculation is to ensure that during normal operation and during anticipated operational occurrences, at least 99.9 percent of all fuel rods in the core do not experience transition boiling if the safety limit is not exceeded.

Based on these considerations, the proposed changes do not involve a significant increase in the probability or consequences of previously analyzed accident.

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

Response: No.

Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation. The SLMCPR is a TS numerical value calculated for two recirculation loop operation and single recirculation loop operation to ensure at least 99.9 percent of all fuel rods in the core do not experience transition boiling if the safety limit is not exceeded. SLMCPR values are calculated using NRC-approved methodology identified in the TS. The proposed SLMCPR values do not involve any new modes of plant operation or any plant modifications and do not directly or indirectly affect the failure modes of any plant systems or components. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The SLMCPR provides a margin of safety by ensuring that at least 99.9 percent of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences if the MCPR Safety Limit is not exceeded. Revision of the SLMCPR values in Technical Specification 2.1.1.2, using NRC-approved methodology, will ensure that the current level of fuel protection is maintained by continuing to ensure that the fuel design safety criterion is met (i.e., that no more than 0.1 percent of the rods are expected to be in boiling transition if the MCPR Safety Limit is not exceeded). Therefore, the proposed amendments do not result in a significant reduction in the margin of safety.

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

Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon St., M/C DEC45A, Charlotte, NC 28202.

NRC Acting Branch Chief: Jeanne A. Dion.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: October 26, 2016. A publicly-available version is in ADAMS, under Accession No. ML16301A150.

Description of amendment request: The proposed amendment would change the Technical Specifications (TS) to revise requirements for unavailable barriers by adding new Limiting Condition for Operation (LCO) 3.0.9. This LCO establishes conditions under which systems would remain operable when required physical barriers are not capable of providing their related support function. This proposed amendment is consistent with NRC-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-427, Revision 2, “Allowance for Non-Technical Specification Barrier Degradation on Supported System OPERABILITLY.” The Notice of Availability of this TS improvement and the model application was published in the Federal Register on October 3, 2006 (71 FR 58444), as part of the consolidated line item improvement process.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee provided an analysis of the issue of no significant hazards consideration (NSHC) by citing the proposed NSHC determination published by the NRC staff in the Federal Register notice referenced above. The proposed NSHC is reproduced below:

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

The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed LCO 3.0.9 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.9. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to an unavailable barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG [Regulatory Guide] 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.9 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP [incremental conditional core damage probability] and ICLERP [incremental conditional large early release probability]) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation. Therefore, this change does not involve a significant reduction in a margin of safety.

Attorney for licensee: William B. Glew, Jr., Associate General Counsel—Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York 10601.

NRC Acting Branch Chief: Douglas A. Broaddus.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: September 12, 2016. A publicly-available version is in ADAMS under Accession No. ML16258A146.

Description of amendment request: The proposed amendment would revise the setpoint for detecting a loss of voltage on the 4.16 kilovolt essential service system (ESS) buses.

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

Response: No.

The proposed change to the 4.16 kV Essential Service System (ESS) bus loss of voltage allowable values allows the protection scheme to function as originally designed. This change will involve alteration of nominal trip setpoints in the field and will also be reflected in revisions to the calibration procedures. The proposed change does not affect the probability or consequences of any accident. Analysis was conducted and demonstrates that the proposed allowable values will allow the normally operating safety related motors to continue to operate without sustaining damage or tripping during the worst-case, non-accident degraded voltage condition for the maximum possible time-delay of 332.3 seconds. Thus, these safety related loads will be available to perform their safety function if a loss-of-coolant accident (LOCA) concurrent with a loss-of offsite power (LOOP) occurs following the degraded voltage condition.

The proposed change does not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration or the plant or the manner in which the plant is operated or maintained. The proposed allowable values ensure that the 4.16 kV distribution system remains connected to the offsite power system when adequate offsite voltage is available and motor starting transients are considered. The emergency diesel generator (EDG) start due to a LOCA signal is not adversely affected by this change. During an actual loss of voltage condition, the loss of voltage time delay will continue to isolate the 4.16 kV distribution system from offsite power before the EDG is ready to assume the emergency loads, which is the limiting time basis for mitigating system responses to the accident. For this reason, the existing loss of power LOCA analysis continues to be valid.

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 involves the revision of 4.16 kV ESS bus loss of voltage allowable values to satisfy existing design requirements. The proposed change does not introduce any changes or mechanisms that create the possibility of a new or different kind of accident. The proposed change does not install any new or different type of equipment, and installed equipment is not being operated in a new or different manner. No new effects on existing equipment are created nor are any new malfunctions introduced.

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

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

Response: No.

The proposed protection voltage allowable values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The EDG start due to a LOCA signal is not adversely affected by this change. During an actual loss of voltage condition, the loss of voltage time delays will continue to isolate the 4.16 kV distribution system from offsite power before the EDG is ready to assume the emergency loads.

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

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

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

NRC Acting Branch Chief: G. Edward Miller.

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

Date of amendment request: September 28, 2016. A publicly-available version is in ADAMS under Package Accession No. ML16277A194.

Description of amendment request: The amendments would revise the BVPS-1 and BVPS-2 Emergency Plan by revising the emergency action level (EAL) schemes to one based on Nuclear Energy Institute (NEI) 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” November 2012 (ADAMS Accession No. ML12326A805). NEI 99-01, Revision 6, was endorsed by the NRC by letter dated March 28, 2013 (ADAMS Accession No. ML12346A463).

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

Response: No.

The proposed changes to BVPS's EAL scheme to adopt the NRC-endorsed guidance of NEI 99-01, Revision 6, do not involve any physical changes to plant systems or equipment. The proposed changes do not alter any of the requirements of the technical specifications. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes do not impact the ability of structures, systems, or components (SSCs) to perform their intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits.

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

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

Response: No.

The proposed changes to BVPS's EAL scheme to adopt the NRC-endorsed guidance of NEI 99-01, Revision 6, do not involve any physical changes to plant systems or equipment. The proposed changes do not involve the addition of any new plant equipment. The proposed changes will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. BVPS functions will continue to be performed as required. The proposed changes do not create any new credible failure mechanisms, malfunctions, or accident initiators.

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

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

Response: No.

The proposed changes to BVPS's EAL scheme to adopt the NRC-endorsed guidance of NEI 99-01, Revision 6, do not involve any physical changes to plant systems or equipment. Margins of safety are unaffected by the proposed changes. There are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed EAL scheme change. The proposed change does not affect the technical specifications. There are no changes to environmental conditions of any of the SSC or the manner in which any SSC is operated. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.

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

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

NRC Acting Branch Chief: Stephen S. Koenick.

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

Date of amendment request: November 1, 2016. A publicly-available version is in ADAMS under Accession No. ML16307A029.

Description of amendment request: The proposed amendment would revise Technical Specification 2.1.1, “Reactor Core [Safety Limits] SLs,” to reduce the reactor steam dome pressure value.

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

Response: No.

Decreasing the reactor steam dome pressure limit in Technical Specification Safety Limits 2.1.1 expands the range of use of the GEXL14 and GEXL17 correlations (applicable to GE14 and GNF2 fuel respectively) and the calculation of the minimum critical power ratio (CPR). The CPR increases during the pressure reduction that occurs during the [Pressure Regulator Failure-Maximum Demand (Open)] PRFO event, so that the initial CPR is the limiting CPR condition during the entire transient. CPR increases during the event relative to the initial CPR value, so fuel cladding integrity is not threatened. Since the change does not involve a modification of any plant hardware, the probability and consequence of the PRFO transient are essentially unchanged.

The proposed change will continue to support the application range of the GEXL correlations applied at PNPP and the calculation of the minimum CPR. The proposed TS revision involves no significant changes to the operation of any systems or components in normal, accident or transient operating conditions.

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

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

Response: No.

The proposed reduction in the reactor steam dome pressure limit in Technical Specification Safety Limits 2.1.1 from 785 psig to 686 psig is a change based on NRC approved documents that permit a wider range of applicability for the GEXL critical power correlations for both GE14 and GNF2 fuel types in the reactor core. This change does not involve changes to the plant hardware or its operating characteristics. There are no changes in the method by which any plant systems perform a safety function, nor are there any changes in the methods governing normal plant operation. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. As a result, no new failure modes are being introduced.

Therefore, the 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 margin of safety is established through the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. Evaluation of the 10 CFR part 21 condition by GE determined that, since the critical power ratio improves during the PRFO transient, there is no impact on the fuel safety margin, and there is no challenge to fuel cladding integrity. The proposed changes do not change the requirements governing operation or the availability of safety equipment assumed to operate to preserve the margin of safety.

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

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

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: October 27, 2016. A publicly-available version is in ADAMS under Accession No. ML16302A055.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.8.3, “Diesel Fuel Oil, Lube Oil, and Starting Air,” by relocating the current stored diesel fuel oil and lube oil numerical volume requirements from the TS to the TS Bases. The proposed changes are consistent with Technical Specifications Task Force Traveler TSTF-501, Revision 1, “Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control.”

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

Response: No

The proposed change relocates the volume of diesel fuel oil and lube oil required to support 7-day operation of each onsite diesel generator, and the volume equivalent to a 6-day supply, to licensee control. The specific volume of fuel oil equivalent to a 7 and 6-day supply is calculated using the NRC-approved methodology described in Regulatory Guide 1.137, Revision 1, “Fuel-Oil Systems for Standby Diesel Generators” and ANSI-N195 1976, “Fuel Oil Systems for Standby Diesel-Generators.” The specific volume of lube oil equivalent to a 7-day and 6-day supply is based on the diesel generator manufacturer's consumption values for the run time of the diesel generator. Because the requirement to maintain a 7-day supply of diesel fuel oil and lube oil is not changed and is consistent with the assumptions in the accident analyses, and the actions taken when the volume of fuel oil and lube oil are less than a 6-day supply have not changed, neither the probability nor the consequences of any accident previously evaluated will be affected.

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

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

Response: No

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 change does not alter assumptions made in the safety analysis but ensures that the diesel generator operates as assumed in the accident analysis. The proposed change is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No

The proposed change relocates the volume of diesel fuel oil and lube oil required to support 7-day operation of each onsite diesel generator, and the volume equivalent to a 6-day supply, to licensee control. As the bases for the existing limits on diesel fuel oil, and lube oil are not changed, no change is made to the accident analysis assumptions and no margin of safety is reduced as part of this change. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: October 7, 2016. A publicly-available version is in ADAMS under Accession No. ML16281A139.

Description of amendment request: The amendment would revise the Hope Creek Technical Specifications by incorporating Nuclear Energy Institute (NEI) topical report 94-01, Revision 3-A, and the conditions and limitations specified in NEI topical report 94-01, Revision 2-A, as the implementation document for the Hope Creek performance-based containment leakage rate testing program. Based on guidance in NEI 94-01, Revision 3-A, the proposed change would allow the Hope Creek Type A Test (Integrated Leak Rate Test, or ILRT) frequency to be extended from 10 to 15 years, and the Type C Tests (Local Leak Rate Tests, or LLRTs) frequency to be extended from 60 to 75 months. In addition, the amendment would delete a one-time extension of the test frequencies previously granted in License Amendment No. 147.

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 revision of the Hope Creek Generating Station (HCGS) Technical Specification (TS) 6.8.4.f, Primary Containment Leakage Rate Testing Program, to allow the extension of the HCGS Type A containment integrated leakage rate test interval to 15 years, and the extension of the Type C local leakage rate 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 existing 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 internal events accident sequences, is 5.15E-03 person-rem/yr (0.01%) using the Electric Power Research Institute (EPRI) guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose. The change in dose risk drops to 1.38E-03 person-rem/yr (<0.01%) when using the EPRI Expert Elicitation methodology. 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,” dated January 1995, 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 HCGS 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 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 test interval extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes an exception previously granted in amendment 147 to allow a one-time extension of the ILRT test frequency for HCGS. 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 HCGS 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 TS 6.8.4.f, “Primary Containment Leakage Rate Testing Program,” involves the extension of the HCGS 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 modification 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 for HCGS.

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

Response: No.

The proposed amendment to TS 6.8.4.f involves the extension of the HCGS 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 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 the extension of the interval between Type A containment leak rate tests and Type C tests for HCGS. 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 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 Types 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 an exception previously granted in amendment 147 to allow a one-time extension of the ILRT test frequency for HCGS. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how HCGS 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: Jeffrie J. Keenan, PSEG Nuclear LLC—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Acting Branch Chief: Stephen S. Koenick.

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

Date of amendment request: September 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16273A557.

Description of amendment request: The changes would amend Combined License Nos. NPF-93 and NPF-94 for the VCSNS, Units 2 and 3, respectively. The amendments propose changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from a plant-specific Design Control Document Tier 2 figure and a Combined Operating License (COL) Appendix C table. Specifically, the proposed departures consist of changes to plant-specific UFSAR Figure 9.3.6-1, Sheet 2 of 2, and COL Appendix C, Table 2.3.2-4, related to the configuration of the boric acid storage tank (BAST) suction point. The change also aligns the Tier 1 Chemical and Volume Control System (CVS) makeup flow rate with previously approved Tier 2 information.

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

Response: No.

The proposed changes alter the BAST suction point by relocating the common inlet/outlet line from the bottom of the tank to the side of the tank and to align the Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for the maximum CVS flow to the Reactor Coolant System (RCS) with the previously approved Tier 2 descriptions and analyses. No change is made to the minimum required volume of the BAST, the included concentrations, or the overall operation of the system. The proposed changes do not alter any safety related functions, and the analyses previously performed on the potential for an inadvertent dilution event are not affected. Consequently, there is no change to an accident initiator in the UFSAR and accordingly, there is no change to the probabilities of accident previously evaluated. The radioactive source terms and release paths are unchanged, thus the radiological releases in the UFSAR accident analyses are not affected.

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 change to alter the BAST suction point affects only nonsafety-related equipment, reducing the possibility for leaks in the BAST. The basic requirements, including the applicable codes and standards, for the configuration of the BAST are unchanged. In addition, the change to the ITAAC verified CVS makeup flow does not alter the design of the CVS, which is currently limited in the design to 175 gallons per minute of flow. The change to the ITAAC aligns the test with the Tier 2 requirement. Consequently, because the BAST codes and standards are unchanged and the CVS is otherwise unchanged, there is no effect on accidents previously evaluated in the UFSAR.

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

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

Response: No.

The proposed change to the BAST piping configuration and to the CVS makeup flow ITAAC does not alter any safety-related equipment, applicable design codes, code compliance, design function, or safety analysis. Consequently, no safety analysis or design basis acceptance limit is challenged or exceeded by the proposed changes, and thus, the margin of safety is not reduced.

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 Branch Chief: Michael T. Markley.

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

Date of amendment request: September 2, 2016. A publicly-available version is in ADAMS under Accession No. ML16246A214.

Description of amendment request: The amendment request proposes changes to a plant-specific Tier 1 (and combined license Appendix C) table and the Updated Final Safety Analysis Report (UFSAR) tables to clarify the flow area for the Automatic Depressurization System (ADS) fourth stage squib valves and to reduce the minimum effective flow area for the second and third stage ADS control valves. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures.

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

Response: No.

The proposed changes do not adversely affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) accident initiator or initiating sequence of events. The proposed changes do not adversely affect the physical design and operation of the second and third stage ADS control valves and fourth stage ADS squib valves, including as-installed inspections, testing, and maintenance requirements, as described in the UFSAR. Therefore, the operation of the second and third stage ADS control valves and fourth stage ADS squib valves is not adversely affected. Inadvertent operation or failure of the second and third stage ADS control valves and fourth stage ADS squib valves are considered as accident initiators or part of an initiating sequence of events for an accident previously evaluated. However, the proposed changes do not adversely affect the probability of inadvertent operation or failure, nor the consequences of such accident precursor sequences. Therefore, the probabilities of the accidents previously evaluated in the UFSAR are not affected.

The proposed changes do not adversely affect the ability of the second and third stage ADS control valves and fourth stage ADS squib valves to perform their design functions. The designs of the second and third stage ADS control valves and fourth stage ADS squib valves continue to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the proposed changes maintain the capabilities of the second and third stage ADS control valves and fourth stage ADS squib valves to mitigate the consequences of an accident and to meet the applicable regulatory acceptance criteria. The proposed changes do not adversely affect the prevention and mitigation of other abnormal events, e.g., anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR are not affected.

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 do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes do not adversely affect the physical design and operation of the second and third stage ADS control valves and fourth stage ADS squib valves, including as-installed inspections, testing, and maintenance requirements, as described in the UFSAR. Therefore, the operation of the second and third stage ADS control valves and fourth stage ADS squib valves is not adversely affected. These proposed changes do not adversely affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that results in significant fuel cladding failures.

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 maintain existing safety margins. The proposed changes maintain the capabilities of the second and third stage ADS control valves and fourth stage ADS squib valves to perform their design functions. The proposed changes maintain existing safety margin through continued application of the existing requirements of the UFSAR, while updating the acceptance criteria for verifying the design features necessary to confirm the second and third stage ADS control valves and fourth stage ADS squib valves perform the design functions required to meet the existing safety margins in the safety analyses. Therefore, the proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced. Therefore, the requested amendment does not involve a significant reduction in a margin of safety.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Inc., (SNC) Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendment request: October 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16285A354.

Description of amendment request: SNC stated that the current Technical Specification (TS) 3.8.9, “Distribution Systems—Operating,” contains a conservative Completion Time of 8 hours for an inoperable 600 Volt alternating current (AC) load center (LC) 1-2R. The proposed change would add new Action Conditions to TS 3.8.9 and include appropriate Required Actions and associated Completion Times for LC 1-2R.

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

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

Response: No.

The proposed amendment revises the TS requirements to include an appropriate Condition, Required Actions and associated Completion Times to address an inoperable 600 Volt AC LC 1-2R that is required to be operable by TS 3.8.9 “Distribution Systems—Operating.”

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The 600V LC are not a precursor to any accident previously evaluated. The proposed changes do 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 changes do not adversely affect the ability of structures, systems and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The LC 1-2R provides power to equipment that may be used to mitigate the consequences of accidents previously evaluated. The proposed change to TS 3.8.9, “Distribution Systems—Operating” provides assurance that the requirements of the TS appropriately address all the equipment that is required to be operable by TS 3.8.9. Thus, the proposed change does not affect the probability or the consequences of any accident previously evaluated.

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

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

Response: No.

The proposed amendment revises the TS to include an appropriate Condition, Required Actions, and associated Completion Times to address inoperable 600 Volt AC LC 1-2R that is required to be operable by TS 3.8.9 “Distribution Systems- Operating.”

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change to the TS assures that the TS appropriately addresses all the equipment required to be operable to support the electrical distribution system. Thus, the proposed change does not adversely affect the design function or operation of any structures, systems, and components important to safety.

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

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

Response: No.

The proposed amendment revises the TS requirements to include an appropriate Condition, Required Actions, and associated Completion Times to address inoperable 600 Volt AC LC 1-2R that is required to be operable by TS 3.8.9 “Distribution Systems—Operating.”

The proposed change to TS 3.8.9 “Distribution Systems—Operating” provides assurance that all the requirements of the TS are appropriately addressed in the Action Conditions. The proposed change serves to make the TS more complete and appropriate for all the equipment required to be operable to support the electrical distribution system. Thus, the proposed change does not involve a change in the margin of safety.

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

Attorney for licensee: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35201.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: October 26, 2016. A publicly-available version is in ADAMS under Accession No. ML16300A325.

Description of amendment request: The proposed changes would amend Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4, respectively. The amendments propose changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from a plant-specific Design Control Document Tier 2 figure and a Combined Operating License (COL) Appendix C table. Specifically, the proposed departures consist of changes to plant-specific UFSAR Figure 9.3.6-1, Sheet 2 of 2, and COL Appendix C, Table 2.3.2-4, related to the configuration of the boric acid storage tank (BAST) suction point. The changes also align the Tier 1 chemical and volume control system (CVS) makeup flow rate with previously approved Tier 2 information.

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

Response: No.

The proposed changes alter the BAST suction point by relocating the common inlet/outlet line from the bottom of the tank to the side of the tank and aligns the Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for the maximum CVS flow to the reactor coolant system (RCS) with the previously approved Tier 2 descriptions and analyses. No change is made to the minimum required volume of the BAST, the included concentrations, or the overall operation of the system. The proposed changes do not alter any safety-related functions, and the analyses previously performed on the potential for an inadvertent dilution event are not affected. Consequently, there is no change to an accident initiator in the UFSAR and accordingly, there is no change to the probabilities of accident previously evaluated. The radioactive source terms and release paths are unchanged, thus the radiological releases in the UFSAR accident analyses are not affected.

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.

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

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

Response: No.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety from any accident previously evaluated.

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: August 31, 2016. A publicly-available version is in ADAMS under Accession No. ML16244A253.

Description of amendment request: The amendment request proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document Tier 2 information and a combined license (COL) License Condition which references one of the proposed changes. Additionally, the proposed changes to the UFSAR eliminate pressurizer spray line monitoring during pressurizer surge line first plant only testing. In addition, these proposed changes correct inconsistencies in testing purpose, testing duration, and the ability to leave equipment in place following the data collection period. These changes involve material which is specifically referenced in Section 2.D.(2) of the COLs. This submittal requests approval of the license amendment necessary to implement these changes.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below with Nuclear Regulatory Commission (NRC) staff edits in square brackets:

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

Response: No.

The design functions of the [reactor coolant system (RCS)] include providing an effective reactor coolant pressure boundary. The proposed changes for removing the requirement to install temporary instrumentation on the pressurizer spray line during the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge l ine valve leakage requirement do not impact the existing design requirements for the RCS. These changes are acceptable as they are consistent with the commitments made for the pressurizer surge line monitoring program for the first plant only, and do not adversely affect the capability of the pressurizer surge line and pressurizer spray lines to perform the required reactor coolant pressure boundary design functions.

These proposed changes to the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement as described in the current licensing basis do not have an adverse effect on any of the design functions of the systems. The proposed changes do not affect the support, design, or operation of mechanical and fluid systems required to mitigate the consequences of an accident. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the proposed changes create any new accident precursors.

Therefore, the requested 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 for removing the requirement to install temporary instrumentation on the pressurizer spray line during the monitoring of the pressurizer surge line for thermal stratification and thermal cycling during hot functional testing and during the first fuel cycle for the first plant only, proposed changes to parameter retention requirements, and proposed change to remove the pressurizer spray and surge line valve leakage requirement as described in the current licensing basis maintain the required design functions, and are consistent with other Updated Final Safety Analysis Report (UFSAR) information. The proposed changes do not adversely affect the design requirements for the RCS, including the pressurizer surge line and pressurizer spray lines. The proposed changes do not adversely affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is adversely affected by the proposed changes.

Therefore, the requested 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.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: October 14, 2016. Publicly-available version is in ADAMS under Accession No. ML16288A810.

Description of amendment request: The requested amendment requires a change to the Combined License (COL) Appendix A, as well as plant-specific Tier 2, Tier 2 *, and COL Appendix C (and corresponding plant-specific Tier 1). The proposed changes would revise the licenses basis documents to add design detail to the automatic depressurization system (ADS) blocking device and to add the blocking device to the design of the in-containment refueling water storage tank injection squib valves actuation logic. An exemption request relating to the proposed changes to the AP1000 Design Control Document Tier 1 is included with the request.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below with Nuclear Regulatory Commission (NRC) staff edits in square brackets:

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

Response: No.

The AP1000 accident analysis previously evaluated a loss of coolant accident caused by an inadvertent ADS valve actuation. Adding design detail to the ADS blocking device, and applying the blocking device to the [in-containment refueling water storage tank (IRWST)] injection valves, does not impact this analysis. Using a blocking device on the ADS and IRWST injection valves is a design feature which further minimizes the probability of a loss of coolant accident caused by a spurious valve actuation. Furthermore, because the blocking device is designed to prevent a spurious valve actuation due to a software [common cause failure (CCF)] and does not adversely impact any existing design feature, it does not involve a significant increase in the probability of an accident previously evaluated.

The proposed amendment does not affect the prevention and mitigation of abnormal events, e.g., accidents, anticipated operation occurrences, earthquakes, floods, turbine missiles, and fires or their safety or design analyses. This change does not involve containment of radioactive isotopes or any adverse effect on a fission product barrier. There is no impact on previously evaluated accidents source terms. The [protection and safety monitoring system (PMS)] is still able to actuate ADS and IRWST injection valves for plant conditions which require their actuation. Therefore, the proposed amendment does not involve a significant increase in the consequences of an accident previously evaluated

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

Response: No.

The proposed changes do not involve a new failure mechanism or malfunction, which affects an SSC accident initiator, or interface with any [structure, system, or component (SSC)] accident initiator or initiating sequence of events considered in the design and licensing bases. There is no adverse effect on radioisotope barriers or the release of radioactive materials. The proposed amendment does not adversely affect any accident, including the possibility of creating a new or different kind of accident from any accident previously evaluated. Therefore, the proposed changes do not create the possibility of a new or different type of accident from any accident previously evaluated.

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

Response: No.

The blocking device is independent of PMS processor hardware and software. It is designed to allow for ADS and IRWST injection actuations when the plant parameters indicate an actual [loss of coolant accident (LOCA)] event. Therefore, the ADS and IRWST are still able to perform their safety functions when required. A postulated failure of a blocking device which would prevent necessary ADS and IRWST injection valve opening would be detected by the proposed periodic surveillance testing within the [Technical Specifications (TS)]. Failure of the ADS actuation or IRWST injection valve opening in a division could also result from concurrent failure of the two Core Makeup Tanks (CMTs) level sensors in one division, with both sensors reading above the blocking setpoint. Failures of the level sensors would be immediately detected due to the deviations in redundant measurements. Furthermore, the proposed TS actions require that the four divisions of blocking devices be capable of automatically unblocking for each CMT. In addition, the TS require that the blocking devices be unblocked in plant modes which allow for the operability of less than two CMTs.

The blocking device will continue to comply with the existing [Updated Final Safety Analysis Report (UFSAR)] regulatory requirements and industry standards. The proposed changes would not affect any safety-related design code, function, design analysis, safety analysis input or result, or existing design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes. Therefore the proposed amendment does not involve a significant reduction in a margin of safety.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

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

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

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

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2 (MPS2), New London County, Connecticut

Date of amendment request: January 26, 2016, as supplemented by letter dated July 14, 2016.

Brief description of amendment: The amendment revised the MPS2 licensing basis to change the spent fuel pool heat load analysis description contained in the Final Safety Analysis Report.

Date of issuance: November 29, 2016.

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

Amendment No.: 330. A publicly-available version is in ADAMS under Accession No. ML16277A680; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-65: Amendment revised the Renewed Operating License.

Date of initial notice in Federal Register: May 24, 2016 (81 FR 32804). The supplemental letter dated July 14, 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 November 29, 2016.

No significant hazards consideration comments received: No.

Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal River Nuclear Generating Plant, Unit 3 (CR-3), Citrus County, Florida

Date of application for amendment: August 27, 2015, as supplemented by letters dated March 2, 2016, and July 14, 2016.

Brief description of amendment: The amendment approves the CR-3 Permanently Defueled Emergency Plan, and Permanently Defueled Emergency Action Level Bases Manual, for the Independent Spent Fuel Storage Installation.

Date of issuance: December 5, 2016.

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

Amendment No.: 252. A publicly-available version is in ADAMS under Accession No. ML16244A102; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. DPR-72: This amendment revises the License.

Date of initial notice in Federal Register: November 10, 2015 (80 FR 69711).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 5, 2016.

No significant hazards consideration comments received: No.

Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: November 2, 2015, as supplemented by letters dated December 22, 2015; and March 31, May 9, and September 14, 2016.

Brief description of amendment: The amendment revised the reactor coolant system (RCS) pressure and temperature limits by replacing Technical Specification (TS) Section 3.4.3, “RCS Pressure and Temperature (P/T) Limits,” Figures 3.4.3-1 and 3.4.3-2, with figures that are applicable up to 50 effective full power years (EFPYs).

Date of issuance: November 22, 2016.

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

Amendment No.: 248. A publicly-available version is in ADAMS under Accession No. ML16285A404; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-23: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: March 1, 2016 (81 FR 10678). The supplemental letters dated March 31, May 9, and September 14, 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 November 22, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 18, 2015, as supplemented by letters dated September 29, 2015, February 5, 2016, April 28, 2016, and May 19, 2016.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.” Additionally, the change added a new program, the Surveillance Frequency Control Program, to TS Section 6, “Administrative Controls.”

Date of issuance: November 29, 2016.

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

Amendment No.: 154. A publicly-available version is in ADAMS under Accession No. ML16200A285; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment.

Renewed Facility Operating License No. NPF-63: Amendment revised the Facility Operating License and TSs.

Date of initial notice in Federal Register: The NRC staff initially made a proposed determination that the amendment request dated August 18, 2015, as supplemented by letter dated September 29, 2015, involved no significant hazards consideration (NSHC) (December 8, 2015, 80 FR 76319). By letters dated February 5, 2016, and April 28, 2016, the licensee provided clarifying information that did not expand the scope of the application and did not change the NRC staff's original proposed NSHC determination, as published in the Federal Register (FR) on December 8, 2015 (80 FR 76319). Subsequently, by letter dated May 19, 2016, the licensee supplemented its amendment request with a proposed change that expanded the scope of the request. Therefore, the NRC published a second proposed NSHC determination in the FR on July 15, 2016 (81 FR 46118), which superseded the notice dated December 8, 2015 (80 FR 76319).

The Commission's related evaluation of the amendment is contained in an SE dated November 29, 2016.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. STN 50-457, Braidwood Station, Unit No. 2, Will County, Illinois

Date of application for amendment: September 30, 2016 (ADAMS Accession No. ML16274A474), as supplemented by letters dated October 26 and 28, 2016, and November 14, 2016 (ADAMS Accession Nos. ML16301A073, ML16302A468, and ML16319A397).

Brief description of amendment: The amendment allows a one-time extension from 72 hours to 200 hours of the technical specification completion time associated with the 2A service water (SX) pump in support of maintenance activities.

Date of issuance: November 23, 2016.

Effective date: As of the date of issuance and shall be implemented prior to the 2A SX pump work window.

Amendment No.: 191. A publicly-available version is in ADAMS under Accession No. ML16315A302; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No NPF-77: The amendment revises the Technical Specifications and License

Date of initial notice in Federal Register: October 21, 2016 (81 FR 72838).

The October 26 and 28, 2016 supplements, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

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

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249. Dresden Nuclear Power Station (DNPS), Unit Nos. 2 and 3, Grundy County, Illinois

Date of application for amendment: December 14, 2015, as supplemented by letter dated

June 30, 2016.

Brief description of amendments: The amendment revises the technical specification (TS) for DNPS, Units 2 and 3, standby or emergency diesel generator (EDG) fuel oil day tank volume as described in TS 3.81, “AC Sources—Operating,” surveillance requirement (SR) 3.8.1.4, from the current value of greater than or equal to (>) 205 gallons to >245 gallons. Raising the EDG fuel oil day tank volume requirement will assure that each EDG can operate for one hour at the maximum allowable operating conditions. The licensee has identified this issue as a non-conservative Technical Specification and administrative controls are currently in-place to assure sufficient fuel oil is available in each fuel oil day tank.

Date of issuance: November 30, 2016.

Effective date: As of the date of issuance and shall be implemented no later than 60 days from the date of issuance.

Amendment Nos.: 252 and 245. A publicly-available version is in ADAMS under Accession No. ML16305A212; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License Nos. DPR-19 and DPR-25: Amendment revises the Renewed Facility Operating Licenses and Technical Specification.

Date of initial notice in Federal Register: March 1, 2016 (81 FR 10680). The supplemental letter dated June 30, 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 November 30, 2016.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: January 15, 2016, as supplemented by letters dated April 19, 2016; May 9, 2016; and June 21, 2016.

Brief description of amendments: The amendments reduce the reactor vessel steam dome pressure specified in the technical specifications (TSs) for the reactor core safety limits. The amendments also revise the setpoint and allowable value for the main steam line low pressure isolation function in the TSs.

Date of issuance: November 21, 2016.

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

Amendment Nos.: 222 and 183. A publicly-available version is in ADAMS under Accession No. ML16272A319; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-39 and NPF-85: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: March 15, 2016 (81 FR 13842). The supplemental letters dated April 19, 2016; May 9, 2016; and June 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 amendments is contained in a Safety Evaluation dated November 21, 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: February 23, 2016.

Brief description of amendment: The amendment revised the Technical Specifications.

Date of issuance: November 22, 2016.

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

Amendment No.: 159. A publicly-available version is in ADAMS under Accession No. ML16281A596; documents related to this amendment is listed in the Safety Evaluation enclosed with this amendment.

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

Date of initial notice in Federal Register: May 10, 2016, (81 FR 28897).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 22, 2016.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey

Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit 1 (NMP1), Oswego County, New York

Date amendment request: August 1, 2016.

Brief description of amendments: The amendments would revise OCNGS's Technical

Specification (TS) Section 2.1, “Safety Limit—Fuel Cladding Integrity,” and NMP1's TS Section 2.1.1, “Fuel Cladding Integrity,” to reduce the steam dome pressure.

Date of issuance: November 29, 2016.

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

Amendment Nos.: 289 for OCNGS and 225 for NMP1. A publicly-available version is in ADAMS under Accession No. ML16256A567; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-16 and DPR-63: Amendments revised the Licenses and Technical Specifications.

Date of initial notice in Federal Register: September 27, 2016 (81 FR 66307).

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

No significant hazards consideration comments received: No.

Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: April 31, 2016, as supplemented by a letter dated August 11, 2016.

Brief description of amendments: The amendments revised Appendix B (Environmental Protection Plan, Section 4.2) of the renewed operating licenses to reflect the “currently applicable” Biological Opinion issued by the National Marine Fisheries Service March 24, 2016.

Date of issuance: December 5, 2016.

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

Amendment Nos.: 236 and 186. A publicly-available version is in ADAMS under Accession No. ML16251A128; documents related to these amendments are listed in the safety evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the Renewed Facility Operating Licenses and Appendix B.

Date of initial notice in Federal Register: June 7, 2016 (81 FR 36621). The supplemental letter dated August 11, 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 December 5, 2016.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company and South Carolina Public Service Authority, 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: December 17, 2015 as supplemented January 11, 2016 and March 16, 2016.

Description of amendment: The amendment authorizes changes to the VCSNS, Units 2 and 3, Updated Final Safety Analysis Report Tier 2* information as well as a change to a license condition to, in part, revise the Wall 11 structure by modifying openings, changing reinforcement detailing, clarifying the classification of building structures for high-energy line break events, crediting the north wall of the Turbine Building first bay wall as a high energy line break barrier and associated missile barriers for protection of Wall 11 from tornado missiles.

Date of issuance: May 31, 2016.

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

Amendment No.: 48. A publicly-available version is in ADAMS under Accession No. ML16109A298; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: February 2, 2016 (81 FR 5499). The supplemental letters dated January 11, 2016, and March 16, 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 the Safety Evaluation dated May 31, 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: May 18, 2016.

Description of amendment: The amendment authorizes changes to the VEGP Units 3 and 4 listed minimum volume of the passive core cooling system core makeup tanks (CMT) reflected in Appendix A, Technical Specifications and the Updated Final Safety Analysis Report of the VEGP Units 3 and 4 Combined Licenses. Specifically, the changes reflect a correction to align licensing documents to reflect the CMT volume given in the VEGP Combined License Tier 1 as 2487 cubic feet is based on and supported by a small-break loss-of-coolant accident safety analysis.

Date of issuance: September 15, 2016.

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

Amendment No.: 53. A publicly-available version is in ADAMS under Accession No. ML16216A394; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: July 5, 2016 (81 FR 43646).

The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated September 15, 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: January 29, 2016, and supplemented by letter dated April 8, 2016.

Description of amendment: The amendment authorizes changes to the VEGP, Units 3 and 4, Updated Final Safety Analysis Report in the form of departures from the incorporated plant specific Design Control Document Tier 2* and Tier 2 information. The changes are also approved in plant-specific technical specifications. The changes incorporate information in WCAP-17524-P-A, Revision 1, “AP1000 Core Reference Report,” which was approved by the U.S. Nuclear Regulatory Commission on February 19, 2015.

Date of issuance: August 19, 2016.

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

Amendment No.: 52. A publicly-available version is in ADAMS under Accession No. ML16201A435; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: March 29, 2016 (81 FR 17501). The supplemental letter dated April 8, 2016, provided additional information that clarified the application, did not expand the scope of the application request 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 the Safety Evaluation dated August 19, 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: December 22, 2015, and supplemented by letters dated May 9, 2016, and May 27, 2016.

Description of amendment: The amendment authorizes changes to the VEGP, Units 3 and 4, Updated Final Safety Analysis Report in the form of departures from the incorporated plant- specific Design Control Document Tier 2* and Tier 2 information with respect to proposed changes to the design of auxiliary building Wall 11, and other changes to the licensing basis for use of seismic Category II structures. It also involves a change to a license condition.

Date of issuance: August 3, 2016.

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

Amendment No.: 51. A publicly-available version is in ADAMS under Accession No. ML16201A298; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: February 16, 2016 (81 FR 7835). The supplemental letters dated May 9, 2016, and May 27, 2016, provided additional information that clarified the application, did not expand the scope of the application request 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 the Safety Evaluation dated August 3, 2016.

No significant hazards consideration comments received: No.

III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

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 for the amendment 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.

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual notice of consideration of issuance of amendment, proposed no significant hazards consideration determination, and opportunity for a hearing.

For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 51.12(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 application for amendment, (2) the amendment to Facility Operating License or Combined License, as applicable, 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.

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

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment.

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).

B. Electronic Submissions (E-Filing)

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition (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.

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC Electronic Filing 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 Electronic Filing Help Desk is available between 9 a.m. and 7 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 stating why there is good cause for not filing electronically and 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, 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.

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

Date of amendment request: November 26, 2016.

Brief description of amendment: The amendment revised the High Pressure Core Spray system and Reactor Core Isolation Cooling system actuation instrumentation technical specifications by adding a footnote indicating that the injection functions of Drywell Pressure-High and Manual Initiation are not required to be operable under low reactor pressure conditions.

Date of issuance: November 29, 2016.

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

Amendment No.: 160. A publicly-available version is in ADAMS under Accession No. ML16333A000; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Public comments requested as to proposed no significant hazards consideration (NSHC): No.

The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a Safety Evaluation dated November 29, 2016.

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

NRC Acting Branch Chief: Douglas Pickett.

Dated at Rockville, Maryland, this 8 day of December, 2016.

For the Nuclear Regulatory Commission.

George A. Wilson, Deputy,

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

[FR Doc. 2016-30438 Filed 12-19-16; 8:45 am]

BILLING CODE 7590-01-P