2017-05990. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

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 February 28, 2017 to March 13, 2017. The last biweekly notice was published on March 14, 2017.

DATES:

Comments must be filed by April 27, 2017. A request for a hearing must be filed by May 30, 2017.

ADDRESSES:

You may submit comments by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0080. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail]. 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:

Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: [phone]; email: [e-mail].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2017-0080, 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-2017-0080.
• 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 [e-mail]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the 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-2017-0080, 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 posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

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

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, 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 petition for leave 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. 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/​. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued.

As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (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.

In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in 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 must 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 the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must 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 the requirements at 10 CFR 2.309(f) 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. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures.

Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the 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). The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document.

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 establish 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 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 the 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 May 30, 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. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c).

If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at 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 his or her 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 and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate 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. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC's Web site at http://www.nrc.gov/​site-help/​e-submittals.html. 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 [e-mail], or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (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. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must 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 document is 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 document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system.

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC's 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 [e-mail], or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 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 adjudicatory documents 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 https://adams.nrc.gov/​ehd,, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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

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

Date of amendment request: February 14, 2017. A publicly available version is in ADAMS under Accession No. ML17053A468.

Description of amendment request: The amendment would revise certain staffing and training requirements, reports, programs, and editorial changes in the Technical Specifications (TS) Table of Contents; Section 1.0, “Definitions”; Section 4.0, “Design Features”; and Section 5.0, “Administrative Controls” that will no longer be applicable once PNPS is permanently defueled.

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

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

Response: No.

The proposed amendment would not take effect until PNPS has permanently ceased operation and entered a permanently defueled condition and the Certified Fuel Handler Training and Retraining Program is approved by the NRC. The proposed amendment would modify the PNPS TS by deleting the portions of the TS that are no longer applicable to a permanently defueled facility, while modifying the other sections to correspond to the permanently defueled condition.

The deletion and modification of provisions of the administrative controls do not directly affect the design of structures, systems, and components (SSCs) necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the spent fuel pool. The changes to the administrative controls are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shutdown and defueled condition of the reactor. Thus, the consequences of an accident previously evaluated are not increased.

In a permanently defueled condition, the only credible accidents are the fuel handling accident (FHA) and those involving radioactive waste systems remaining in service. The probability of occurrence of previously evaluated accidents is not increased, because extended operation in a defueled condition will be the only operation allowed. This mode of operation is bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

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 have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The administrative removal or modifications of the TS that are related only to administration of the facility cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled and PNPS will no longer be authorized to operate the reactor or retain or place fuel in the reactor vessel.

The proposed changes to the PNPS TS do not affect systems credited in the accident analysis for the FHA or radioactive waste system upsets at PNPS. The proposed TS will continue to require proper control and monitoring of safety significant parameters and activities.

The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding and spent fuel cooling). Extended operation in a defueled condition will be the only operation allowed, and it is bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

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

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

Response: No.

Since the 10 CFR part 50 license for PNPS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel once the certifications required by 10 CFR 50.82(a)(1) are docketed, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The only remaining credible accidents are a FHA and those involving radioactive waste systems remaining in service. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact these analyzed conditions.

The proposed changes are limited to those portions of the TS that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the PNPS TS are not credited in the existing accident analysis for the remaining applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated design basis accidents involving the reactor are no longer possible because the reactor will be permanently shutdown and defueled and PNPS will no longer be authorized to operate the reactor or retain or place fuel in the reactor vessel.

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

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

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

NRC Branch Chief: Douglas A. Broaddus.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of amendment request: December 13, 2016, as supplemented by letter dated February 17, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML16348A368 and ML17048A034, respectively.

Description of amendment request: The amendment would revise the NMP2 technical specification (TS) safety limit (SL) to increase the low pressure isolation setpoint allowable value, which will result in earlier main steam line isolation. The revised main steam line low pressure isolation capability and the revised SL are intended to ensure that NMP2 remains within the TS SLs in the event of a pressure regulator failure maximum demand transient.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below, with NRC staff 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 proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because decreasing the reactor dome pressure in TS SL 2.1.1.1 and TS SL 2.1.1.2 for reactor RTP [rated thermal power] ranges and increasing the AV [allowable value] for the Main Steam Line Pressure-Low on TS Table 3.3.6.1-1, Function b, effectively expands the range of applicability for GEXL correlation and the calculation of MCPR [minimum critical power ratio]. The CPR [critical power ratio] rises during the pressure reduction following the scram that terminates the PRFO [pressure regulator failure—maximum demand (open)] transient. The reduction in the reactor dome pressure value in the SL from 785 psig [pounds per square inch gauge] to 700 psia [pounds per square inch absolute] and the increase in the AV from ≥746 psig to ≥814 psig adequately accommodate the pressure reduction during the PRFO transient within the revised TS limit without compromising fuel integrity.

The expanded GEXL correlation range supports NMP2 revised low pressure safety limit of 700 psia. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident or transient operating conditions.

Therefore, the proposed changes do 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 create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed reduction in the reactor dome pressure value in the SL from 785 psig to 700 psia reflects a wider range of applicability for the GEXL correlation which is approved by the NRC for both GE14 currently in NMP2 and GNF2 fuels proposed for NMP2. The proposed changes do not involve physical changes to the plant or its operating characteristics. In addition, the increase in the AV for the MSL [main steam line] low pressure from ≥746 psig to ≥814 psig will result in the MSIV [main steam isolation valve] closure signal initiation at a higher temperature. As a result, no new failure modes are being introduced.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not involve a significant reduction in a margin of safety because 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. The proposed change in reactor dome pressure SLs and the AV for the MSL low pressure ensures the safety margin is maintained, which protects the fuel cladding integrity during steady state operation, normal operational transients, or AOOs [anticipated operational occurrences] such as a depressurization transient, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The proposed changes do not involve physical changes to the plant or its operating characteristics. The reduction in the reactor dome pressure value in the SL from 785 psig to 700 psia and the increase to the AV for the MSL low pressure provides added margin to accommodate the pressure reduction during the PRFO transient within the revised TS limit without compromising fuel integrity.

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

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

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

NRC Acting Branch Chief: Stephen S. Koenick.

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

Date of amendment request: February 20, 2017. A publicly-available version is available in ADAMS under Accession No. ML17051A003.

Description of amendment request: The licensee proposes to delete from the Facility Operating License (FOL) certain license conditions, which impose specific requirements on the decommissioning trust agreement. The licensee proposes to meet the provisions of 10 CFR 50.75(h) for OCNGS.

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

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

Response: No.

The requested changes delete License Conditions 3.F through 3.K pertaining to Decommissioning Trust Agreements currently in the OCNGS FOL. The requested changes are consistent with the types of license amendments [identified] in 10 CFR 50.75(h)(4).

The regulations of 10 CFR 50.75(h)(4) state “Unless otherwise determined by the Commission with regard to a specific application, the Commission has determined that any amendment to the license of a utilization facility that does no more than delete specific license conditions relating to the terms and conditions of decommissioning trust agreements involves “no significant hazard considerations.”

This request involves changes that are administrative in nature. No actual plant equipment or accident analyses will be affected by the proposed changes.

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

2. Does the [p]roposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

This request involves administrative changes to the license that will be consistent with the NRC's regulations at 10 CFR 50.75(h).

No actual plant equipment or accident analyses will be affected by the proposed change and no failure modes not bounded by previously evaluated accidents will be created.

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.

Margin of safety is associated with confidence in the ability of the fission product barriers to limit the level of radiation dose to the public.

This request involves administrative changes to the license that will be consistent with the NRC's regulations at 10 CFR 50.75(h).

No actual plant equipment or accident analyses will be affected by the proposed change. Additionally, the proposed changes will not relax any criteria used to establish safety limits, will not relax any safety systems settings, or will not relax the bases for any limiting conditions of operation.

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

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

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

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: January 30, 2017. A publicly-available version is in ADAMS under Accession No. ML17030A302.

Description of amendment request: The amendments would replace existing Technical Specification (TS) requirements related to “operations with a potential for draining the reactor vessel” (OPDRVs) with new requirements on reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires RPV water level to be greater than the top of active irradiated fuel. The proposed changes are based on TS Task Force (TSTF) Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control.”

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

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.

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 changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements. The event of concern under the current requirements and the proposed changes are an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

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 proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

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

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

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

NRC Branch Chief: James G. Danna.

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: January 23, 2017. A publicly-available version is in ADAMS under Accession No. ML17025A399.

Description of amendment request: The amendments would modify the St. Lucie Plant, Unit Nos. 1 and 2, Technical Specifications (TSs) by limiting the MODE of applicability for the Reactor Protection System (RPS), Startup, and Operating Rate of Change of Power—High, functional unit trip. Additionally, the proposed license amendments add new Limiting Condition for Operation (LCO) 3.0.5 and relatedly modifies LCO 3.0.2, to provide for placing inoperable equipment under administrative control for the purpose of conducting testing required to demonstrate OPERABILITY.

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

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

Response: No.

Limiting the MODE 1 applicability for RPS functional unit, Startup and Operating Rate of Change of Power—High, to Power Range Neutron Flux Power ≤15% of RATED THERMAL POWER, is an administrative change in nature and does not alter the manner in which the functional unit is operated or maintained. The proposed changes do not represent any physical change to plant [structures, systems, and components (SSC(s))], or to procedures established for plant operation. The subject RPS functional unit is not an event initiator nor is it credited in the mitigation of any event or credited in the [probabilistic risk assessment (PRA)]. As such, the initial conditions associated with accidents previously evaluated and plant systems credited for mitigating the consequences of accidents previously evaluated remain unchanged.

The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 and Unit 2 TS and related modification to LCO 3.0.2 is consistent with the guidance provided in NUREG-1432, Volume 1 [ADAMS Accession No. ML12102A165] (Reference 6.1 [of the amendment request]) and thereby has been previously evaluated by the Commission with a determination that the proposed change does not involve a significant hazards consideration.

Therefore, facility operation in accordance with the proposed license amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

Limiting the MODE 1 applicability for the RPS functional unit, Startup and Operating Rate of Change of Power—High, to Power Range Neutron Flux Power ≤ 5% of RATED THERMAL POWER, is an administrative change in nature and does not involve the addition of any plant equipment, methodology or analyses. The proposed changes do not alter the design, configuration, or method of operation of the subject RPS functional unit or of any other SSC. More specifically, the proposed changes neither alter the power rate-of-change trip function nor its ability to bypass and reset as required. The subject RPS functional unit remains capable of performing its design function.

The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 and Unit 2 TS and related modification to LCO 3.0.2 is consistent with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1 [of the amendment request]) and thereby has been previously evaluated by the Commission with a determination that the proposed change does not involve a significant hazards consideration.

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.

Limiting the MODE 1 applicability for RPS functional unit, Startup and Operating Rate of Change of Power—High, to Power Range Neutron Flux Power ≤15% of RATED THERMAL POWER is an administrative change in nature. The proposed changes neither involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings nor do they adversely impact plant operating margins or the reliability of equipment credited in safety analyses.

The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 and Unit 2 TS and related modification to LCO 3.0.2 is consistent with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1 [of the amendment request]) and thereby has been previously evaluated by the Commission with a determination that the proposed change does not involve a significant hazards consideration.

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

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

Attorney for licensee: William S. Blair, Managing Attorney—Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/JB, Juno Beach, FL 33408-0420.

NRC Branch Chief: Benjamin G. Beasley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1 (FCS), Washington County, Nebraska

Date of amendment request: December 16, 2016. A publicly-available version is in ADAMS under Accession No. ML16351A464.

Description of amendment request: The proposed amendment would revise the FCS Emergency Plan and Emergency Action Level (EAL) scheme for the permanently defueled condition. The proposed permanently defueled Emergency Plan and EAL scheme are commensurate with the significantly reduced spectrum of credible accidents that can occur in the permanently defueled condition and are necessary to properly reflect the conditions of the facility while continuing to preserve the effectiveness of the emergency plan.

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

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

Response: No.

The proposed changes to the FCS Emergency Plan and EAL scheme do not impact the function of facility structures, systems, or components. The proposed changes do not affect accident initiators or precursors, nor does it alter design assumptions. The proposed changes do not prevent the ability of the on-shift staff and emergency response organization to perform their intended functions to mitigate the consequences of any accident or event that will be credible in the permanently defueled condition.

The probability of occurrence of previously evaluated accidents is not increased, because most previously analyzed accidents can no longer occur and the probability of the few remaining credible accidents are unaffected by the proposed amendment.

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 reduce the scope of the FCS Emergency Plan and EAL scheme commensurate with the hazards associated with a permanently shutdown and defueled facility. The proposed changes do not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed changes do not result in a change to the way that the equipment or facility is operated resulting in new or different kinds of accident initiators or accident mitigation.

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

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

Response: No.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the FCS Emergency Plan and EAL scheme and do not impact operation of the facility or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed changes do not involve a change in the method of facility operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by the proposed changes. The revised Emergency Plan will continue to provide the necessary response staff.

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

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

Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817.

NRC Branch Chief: Douglas A. Broaddus.

PSEG Nuclear LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek Generating Station (HCGS) and Salem Nuclear Generating Station (SGS), Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: February 13, 2017. A publicly-available version is in ADAMS under Package Accession No. ML17044A346.

Description of amendment request: The amendments would revise the HCGS and SGS, Unit Nos. 1 and 2, emergency action level (EAL) schemes. Specifically, the licensee proposes to adopt the EAL scheme described in Nuclear Energy Institute (NEI) 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors.” NEI 99-01, Revision 6, has been endorsed by the NRC.

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

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

Response: No.

The proposed changes to the HCGS and SGS EALs do not impact the physical function of plant structures, systems or components (SSC) or the manner in which SSCs perform their design function. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are affected by the proposed changes. 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 changes do not involve a physical alteration of the plant (i.e., no new or different types of equipment will be installed or removed) or a change in the method of plant operation. The proposed changes will not introduce failure modes that could result in a new accident, and the changes do not alter assumptions made in the safety analysis. The proposed changes to the HCGS and SGS EALs are not initiators of any accidents. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not impact operation of the plant or its response to transients or accidents. The changes do not affect the Technical Specifications or the operating license. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. The emergency plan will continue to activate an emergency response commensurate with the extent of degradation of plant 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: Jeffrie J. Keenan, PSEG Nuclear LLC—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: James G. Danna.

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: February 15, 2017. A publicly-available version is in ADAMS under Accession No. ML17046A660.

Description of amendment request: The amendment request proposes to revise the licensing basis information to reflect changes to the locations of the hydrogen venting primary openings in the passive core cooling system (PXS) valve/accumulator rooms inside containment. Because this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).

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

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

Response: No.

The proposed revision to the hydrogen venting for the Passive Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and clarification of the venting path definition for PXS Valve/Accumulator Room B (Room 11207) do not affect any safety-related equipment or function. The hydrogen ignition subsystem, including designed hydrogen venting features, is designed to mitigate beyond design basis hydrogen generation in the containment. The hydrogen venting changes do not involve any accident, initiating event or component failure; thus, the probabilities of the accidents previously evaluated are not affected. The modified venting locations and definitions will maintain the hydrogen ignition subsystem designed and analyzed beyond design basis function to maintain containment integrity. The maximum allowable containment leakage rate specified in the Technical Specifications is unchanged, and radiological material release source terms are not affected; thus, the radiological releases in the 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 revision to the hydrogen venting for the Passive Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and clarification of the venting path definition for PXS Valve/Accumulator Room B (Room 11207) will maintain the beyond design basis function of the hydrogen ignition subsystem. The hydrogen venting changes do not impact the hydrogen ignition subsystem's function to maintain containment integrity during beyond design basis accident conditions, and, thus does not introduce any new failure mode. The proposed changes do not create a new fault or sequence of events that could result in a radioactive release. The proposed changes would not affect any safety-related accident mitigating function.

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 revision to the hydrogen venting for the Passive Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and clarification of the venting path definition for PXS Valve/Accumulator Room B (Room 11207) will maintain the beyond design basis function of the hydrogen ignition subsystem. The proposed changes do not have any effect on the ability of safety-related structures, systems, or components to perform their beyond design basis functions. The proposed changes are a result of a low probability, severe accident scenario being evaluated. The revision to this scenario does not result in an increase in the plant risk (frequency and/or consequences). The frequency is low and there is no increase to the consequences because containment integrity is maintained and there is no containment leakage. There is no change to the maximum allowed containment leakage rate (0.10% of containment air weight per day) for the containment vessel. The proposed changes do not affect the ability of the hydrogen igniter subsystem to maintain containment integrity following a beyond design basis accident. The hydrogen igniter subsystem continues to meet the requirements for which it was designed and continues to meet the regulations.

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

The NRC staff has reviewed the licensee's analysis and, based on 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: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514.

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: February 16, 2017. A publicly-available version is in ADAMS under Accession No. ML17047A192.

Description of amendment request: The requested amendment proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) and involves changes to related plant-specific Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C information, to clarify text that currently refers to raceways with an electrical classification (i.e., Class 1E/non-Class 1E). This includes rewording multiple Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) and UFSAR material to clarify that any text referring to Class 1E or non-Class 1E raceways or raceway systems is referring to raceways or raceway systems that route Class 1E or non-Class 1E circuits.

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

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

Response: No.

These proposed changes are for clarification and consistency. No structure, system, or component (SSC) or function is changed within this activity. There is no change to the application of regulatory guides or industry standards to raceways or raceway systems, nor is there a change to how they are designed, fabricated, procured or installed. Raceway systems that route Class 1E circuits will continue to be designated and designed as equipment Class C, safety-related, and seismic Category I structures. The proposal to align the text in COL Appendix C (and plant-specific Tier 1) Section 3.3 with the associated ITAAC is made for clarification and consistency to reduce misinterpretation. The proposal to reword multiple ITAAC in 3.3.00.07 does not change the intent of the ITAAC, nor is the ITAAC scope or closure method impacted.

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.

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 involve a new failure mechanism or malfunction, which affects an SSC accident initiator, or interface with any 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.

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

Response: No.

These proposed changes are for clarification and consistency to reduce misinterpretation. No SSC or function is changed within this activity. There is no change to the application of regulatory guides or industry standards to raceways or raceway systems, nor is there a change to how they are designed, fabricated, procured or installed. Raceway systems that route Class 1E circuits will continue to be designated and designed as Equipment Class C, safety-related, and seismic Category I.

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.

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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: August 30, 2016. A publicly-available version is in ADAMS under Accession No. ML16243A373.

Description of amendment request: The amendment request proposes a change to Updated Final Safety Analysis Report in the form of departures from the incorporated plant-specific Design Control Document (DCD) Tier 2 * information and related changes to the VEGP Units 3 and 4 Combined License (COL) Appendix C (and corresponding plant-specific DCD Tier 1) information.

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, a design certification rule is also requested for the plant-specific Tier 1 material departures. The proposed change is to the thickness of one floor in the auxiliary building located between Column Lines I to J-1 and Column Lines 2 to 4 at Elevation 153′-0″. 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 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 nuclear island structures are to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in the nuclear island. The nuclear island structures are structurally designed to meet seismic Category I requirements as defined in Regulatory Guide 1.29.

The change of the thickness of the floor above the [Component Cooling Water System (CCS)] Valve Room in the auxiliary building meets criteria and requirements of American Concrete Institute (ACI) 349 and American Institute of Steel Construction (AISC) N690 and does not have an adverse impact on the response of the nuclear island structures safe shutdown earthquake ground motions or loads due to anticipated transient or postulated accident conditions. The proposed changes do not impact the support, design, or operation of mechanical and fluid systems. 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 normal operation or postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor does the change described create any new accident precursors.

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 is to revise the thickness of the floor above the CCS Valve Room in the auxiliary building. The proposed changes do not change the design requirements of the nuclear island structures. The proposed changes do not change the design function, support, design, or operation of mechanical and fluid systems. The proposed changes do not result in a new failure mechanism for the nuclear island structures or new accident precursors. As a result, the design function of the nuclear island structures is not adversely affected by the proposed change.

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.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus, no margin of safety is reduced.

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

The NRC staff has reviewed the licensee's analysis and, based on 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: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: January 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17031A446.

Description of amendment request: The requested amendment proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) and to change Combined License Appendix A, Technical Specifications (TS), to modify engineered safety features logic for containment vacuum relief actuation.

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

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

Response: No.

The proposed changes to the UFSAR and TS will include the Containment Pressure—Low automatic reset function for the containment vacuum relief valves manual initiation logic, such that the containment vacuum relief manual actuation will be automatically reset when the containment pressure rises above the Containment Pressure—Low setpoint. This reset allows a containment isolation signal to close the valves when necessary. The Containment Pressure—Low signal is an interlock for the containment vacuum relief manual actuation such that the valves cannot be opened unless the Containment Pressure—Low setpoint has been reached in any two-out-of-four divisions. The modified logic will ensure that the automatic initiation of containment isolation is made available following manual initiation of containment vacuum relief actuation. The analyzed design and function of the Engineered Safety Features Actuation System and its actuated components is not affected. The proposed changes do not adversely affect any safety-related equipment and does not involve any accident, initiating event, or component failure, thus the probabilities of accidents previously evaluated are not affected. The proposed changes do not adversely interface with or adversely affect any system containing radioactivity or affect any radiological material release source term; thus the radiological releases in an accident 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 changes to the UFSAR and TS to include the Containment Pressure—Low manual actuation interlock and automatic reset function for the containment vacuum relief valves manual initiation logic will maintain the Engineered Safety Features Actuation System and Plant Safety and Monitoring System in accordance with the design objectives as licensed. The design of the Class 1E Containment Pressure—Low manual actuation interlock and automatic reset function is required to meet the licensing basis for the Engineered Safety Features Actuation System and Plant Safety and Monitoring System. The changes to the manual initiation logic do not adversely affect the function of any safety-related structure, system, or component, and thus does not introduce a new failure mode. The changes to the containment vacuum relief valves manual initiation logic do not adversely interface with any safety-related equipment or any equipment associated with radioactive material and, thus, do not create a new fault or sequence of events that could result in a new or different kind of accident.

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 changes to the UFSAR and TS to include the Containment Pressure—Low automatic reset function for the containment vacuum relief valves manual initiation logic will maintain the Engineered Safety Features Actuation System and Plant Safety and Monitoring System in accordance with the design objectives as licensed. The changes to the manual initiation logic do not adversely interface with any safety-related equipment or adversely affect any safety-related function. The changes to the containment vacuum relief manual initiation logic continue to comply with existing design codes and regulatory criteria, and do not involve a significant reduction in the margin of safety.

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

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

Attorney for licensee: Mr. 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: March 2, 2017. A publicly-available version is in ADAMS under Accession No. ML17061A747.

Description of amendment request: The requested amendment consist of changes to Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) in combined license (COL) Appendix C, with corresponding changes to the associated plant-specific Tier 1 information, to consolidate a number of ITAAC to improve efficiency of the ITAAC completion and closure process.

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.

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

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

Response: No.

The proposed non-technical change to COL Appendix C will consolidate, relocate and subsume redundant ITAAC in order to improve and create a more efficient process for the ITAAC Closure Notification submittals. No structure, system, or component (SSC) design or function is affected. No design or safety analysis is affected. The proposed changes do not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the 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 COL Appendix C does not affect the design or function of any SSC, but will consolidate, relocate and subsume redundant ITAAC in order to improve efficiency of the ITAAC completion and closure process. The proposed changes would not introduce a new failure mode, fault or sequence of events that could result in a radioactive material release.

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

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

Response: No.

The proposed change to COL Appendix C to consolidate, relocate and subsume redundant ITAAC in order to improve efficiency of the ITAAC completion and closure process is considered non-technical and would not affect any design parameter, function or analysis. There would be no change to an existing design basis, design function, regulatory criterion, or analysis. No safety analysis or design basis acceptance limit/criterion is involved.

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

The NRC staff has reviewed the licensee's analysis and, based on 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: 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: February 22, 2017. A publicly-available version is in ADAMS under Accession No. ML17053A425.

Description of amendment request: The amendment request proposes to revise the licensing basis information to reflect changes to the locations of the hydrogen venting primary openings in the passive core cooling system (PXS) valve/accumulator rooms inside containment. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 Design Control Document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).

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

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

Response: No.

The proposed revision to the hydrogen venting for the Passive Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and clarification of the venting path definition for PXS Valve/Accumulator Room B (Room 11207) do not affect any safety-related equipment or function. The hydrogen ignition subsystem, including designed hydrogen venting features, is designed to mitigate beyond design basis hydrogen generation in the containment. The hydrogen venting changes do not involve any accident, initiating event or component failure; thus, the probabilities of the accidents previously evaluated are not affected. The modified venting locations and definitions will maintain the hydrogen ignition subsystem designed and analyzed beyond design basis function to maintain containment integrity. The maximum allowable containment leakage rate specified in the Technical Specifications is unchanged, and radiological material release source terms are not affected; thus, the radiological releases in the 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 revision to the hydrogen venting for the PXS Valve/Accumulator Room A (Room 11206) and clarification of the venting path definition for PXS Valve/Accumulator Room B (Room 11207) will maintain the beyond design basis function of the hydrogen ignition subsystem. The hydrogen venting changes do not impact the hydrogen ignition subsystem's function to maintain containment integrity during beyond design basis accident conditions, and, thus does not introduce any new failure mode. The proposed changes do not create a new fault or sequence of events that could result in a radioactive release. The proposed changes would not affect any safety-related accident mitigating function.

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 revision to the hydrogen venting for the Passive Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and clarification of the venting path definition for PXS Valve/Accumulator Room B (Room 11207) will maintain the beyond design basis function of the hydrogen ignition subsystem. The proposed changes do not have any effect on the ability of safety-related structures, systems, or components to perform their beyond design basis functions. The proposed changes are a result of a low probability, severe accident scenario being evaluated. The revision to this scenario does not result in an increase in the plant risk (frequency and/or consequences). The frequency is low and there is no increase to the consequences because containment integrity is maintained and there is no containment leakage. There is no change to the maximum allowed containment leakage rate (0.10% of containment air weight per day) for the containment vessel. The proposed changes do not affect the ability of the hydrogen igniter subsystem to maintain containment integrity following a beyond design basis accident. The hydrogen igniter subsystem continues to meet the requirements for which it was designed and continues to meet the regulations.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

Tennessee Valley Authority (TVA), Docket No. 50-391, Watts Bar Nuclear Plant (WBN), Unit 2, Rhea County, Tennessee

Date of amendment request: December 21, 2016. A publicly-available version is in ADAMS under Accession No. ML16356A673.

Description of amendment request: The amendment would revise the containment ice mass limits in WBN, Unit 2, Technical Specification (TS) Surveillance Requirements (SRs) 3.6.11.2 and 3.6.11.3 to be identical to the ice mass limits in the WBN, Unit 1, TS SRs 3.6.11.2 and 3.6.11.3.

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

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

Response: No.

The primary purpose of the ice bed is to provide a large heat sink to limit peak containment pressure in the event of a release of energy from a design basis LOCA [loss-of-coolant accident] or high energy line break (HELB) in containment. The LOCA requires the greatest amount of ice compared to other accident scenarios; therefore, the reduction in ice weight is based on the LOCA analysis. The amount of ice in the bed has no impact on the initiation of an accident, but rather on the mitigation of the accident. The containment integrity analysis shows that the proposed reduced ice weight is sufficient to maintain the peak containment pressure below the containment design pressure, and that the containment heat removal systems function to rapidly reduce the containment pressure and temperature in the event of a LOCA. Therefore, containment integrity is maintained and the consequences of an accident previously evaluated in the WBN dual-unit Updated Final Safety Analysis Report (UFSAR) are not significantly increased. 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 ice condenser serves to limit the peak pressure inside containment following a LOCA. TVA has evaluated the revised containment pressure analysis and determined that sufficient ice would be present to maintain the peak containment pressure below the containment design pressure. Therefore, the reduced ice weight does not create the possibility of an accident that is different than any already evaluated in the WBN dual-unit UFSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed change.

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

Response: No.

The proposed TS ice weight SR limit is based on the conservatism of the WBN Unit 1 WCOBRA/TRAC LOCA M&E [mass and energy] methodology in comparison to the WBN Unit 2 operating conditions. The WBN Unit 1 WCOBRA/TRAC LOCA M&E methodology is modeled on the WBN Unit 1 RSGs [replacement steam generators], which have a greater mass, volume, and stored metal energy than the WBN Unit 2 original model D3 SGs [steam generators]. Additionally, the containment pressure calculations in Section 6.2.1.3.3 of the WBN Unit 1 portion of the WBN dual-unit UFSAR state that the analytical limit for the mass of ice assumed in the WBN Unit 1 ice condenser, in order to limit the maximum containment peak pressure from a LOCA to below the containment design pressure, is 2,260,000 lb. The proposed revised TS SR ice mass limit of 2,404,500 lb [pound] includes additional ice mass to conservatively bound ice bed sublimation effects. Based on TVA's evaluation and the revised containment analysis, TVA considers the reduction of the ice mass limit to be acceptable for satisfying the safety function of the ice condenser for the current SR interval. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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

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

NRC Branch Chief: Benjamin G. Beasley.

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

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

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

Tennessee Valley Authority, Docket No. 50-391 Watts Bar Nuclear Plant, Unit 2, Rhea County, Tennessee

Date of amendment request: November 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16320A161.

Brief description of amendment request: The proposed amendment would revise the Watts Bar Nuclear Plant, Unit 2, Cyber Security Plan Implementation Schedule for Milestone 8 and would revise the associated license condition in the Facility Operating License.

Date of publication of individual notice in Federal Register: January 5, 2017 (82 FR 1370).

Expiration date of individual notice: February 6, 2017 (public comments); March 6, 2017 (hearing requests).

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

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

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

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

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

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

Date of amendment request: March 22, 2016, as supplemented by letter dated August 11, 2016.

Brief description of amendment: The amendment revised Technical Specification (TS) 5.5.12, “Primary Containment Leakage Rate Testing Program,” for the permanent extension of the Type A test interval up to one test in 15 years, as stipulated in Nuclear Energy Institute (NEI) 94-01, Revision 2-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,” October 2008 (ADAMS Accession No. ML100620847). The license amendment request also proposes to increase the containment isolation valves leakage test intervals (i.e., Type C tests) from their current 60 months to 75 months by replacing TS 5.5.12.a. reference to Regulatory Guide 1.163, “Performance-Based Containment Leak-Test Program” (ADAMS Accession No. ML003740058), with a reference to NEI 94-01, Revision 3-A (ADAMS Accession No. ML12221A202), and the conditions and limitations specified in NEI 94-01, Revision 2-A, to implement the performance-based leakage testing program in accordance with title 10 of the Code of Federal Regulations part 50, Appendix J, Option B. The amendment also deletes from TS 5.5.12, text that authorized a one-time extension of the Type A test interval to 2007 and revised paragraph 2.D of the renewed facility operating license to reflect removal of a reference to an exemption from 10 CFR part 50, Appendix J, requirements for testing of containment air locks.

Date of issuance: March 9, 2017.

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

Amendment No.: 205. A publicly-available version is in ADAMS under Accession No. ML16351A460; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-43: Amendment revised the renewed facility operating license and TSs.

Date of initial notice in Federal Register: June 7, 2016 (81 FR 36616). The August 11, 2016 supplement 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 hazard consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 9, 2017.

No significant hazards consideration comments received: No.

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.

Brief description of amendments: The amendments revised Technical Specification Section 2.1.1.2 to change the minimum critical power ratio safety limit.

Date of issuance: March 10, 2017.

Effective date: As of date of issuance and shall be implemented for Unit 1 prior to start-up from the 2018 refueling outage (March 2018) and for Unit 2 prior to start-up from the 2017 refueling outage.

Amendment Nos.: 272 (Unit 1) and 300 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17059D146; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-71 and DPR-62: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: December 20, 2016 (81 FR 92866).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 10, 2017.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 5, 2016, as supplemented by letter dated June 16, 2016.

Brief description of amendments: The amendments would modify the McGuire Nuclear Station, Units 1 and 2, Technical Specifications (TS) by removing footnote (c) from TS Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation,” which is no longer applicable, and by removing an expired footnote from TS 3.8.1, “AC Sources—Operating.”

Date of issuance: March 8, 2017.

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

Amendment Nos.: 293 and 272. A publicly-available version is in ADAMS under Accession No. ML17003A019; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: July 5, 2016 (81 FR 43649). The supplemental letter dated June 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 amendments is contained in a Safety Evaluation dated March 8, 2017.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of application for amendment: May 10, 2016, as supplemented by letters dated May 18, 2016, and January 31, 2017.

Brief description of amendment: The amendment revised the safety function lift and lower setpoint tolerances of the safety/relief valves that are listed in Surveillance Requirements 3.4.3.1 and 3.4.4.1 of the Technical Specifications.

Date of issuance: March 9, 2017.

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

Amendment No.: 240. A publicly-available version is in ADAMS under Accession No. ML17052A125; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-21: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: July 19, 2016 (81 FR 46961). The supplemental letter January 31, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 9, 2017.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 1, Pope County, Arkansas

Date of amendment request: March 25, 2016.

Brief description of amendment: The amendment deleted Technical Specification (TS) 5.5.8, “Inservice Testing Program.” A new defined term, “Inservice Testing Program,” is added to TS Section 1.1, “Definitions.” Also, existing uses of the term “Inservice Testing Program” in the TSs are capitalized throughout to indicate that it is now a defined term. The NRC staff has concluded that the amendment is consistent with Technical Specifications Task Force Traveler TSTF-545, Revision 3, which was made available to the TSTF via NRC letter dated December 11, 2015 (ADAMS Accession No. ML15317A071).

Date of issuance: March 10, 2017.

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

Amendment No.: 257. A publicly-available version is in ADAMS under Accession No. ML16165A423; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-51: The amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: June 7, 2016 (81 FR 36619).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 10, 2017.

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: April 4, 2016.

Brief description of amendments: The amendments revised the technical specification (TS) requirements for the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) system actuation instrumentation. Specifically, the amendments add a footnote to the TSs indicating that the injection functions of drywell pressure-high (HPCI only) and manual initiation (HPCI and RCIC) are not required to be operable under low reactor pressure conditions.

Date of issuance: February 28, 2017.

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

Amendment Nos.: 224 (Unit 1) and 185 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16356A272; 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: June 7, 2016 (81 FR 36620).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 28, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 3, 2017.

Brief description of amendment: The amendment revised the NMP1 licensing basis related to alternative source term analysis in the updated final safety analysis report (UFSAR) to allow the use of the release fractions listed in Tables 1 and 3 of NRC Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” July 2000 (ADAMS Accession No. ML003716792), for partial length fuel rods (PLRs) that are operating above the peak burnup limit for the remainder of the current operating cycle. In addition, the proposed change revised the NMP1 licensing basis to allow movement of irradiated fuel bundles containing PLRs that have been in operation above 62,000 megawatt days per metric tons of uranium (MWD/MTU).

Date of issuance: March 9, 2017.

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

Amendment No.: 226. A publicly-available version is in ADAMS under Accession No. ML17055A451; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-63: Amendment revised the licensing basis related to alternative source term analysis in the UFSAR.

Date of initial notice in Federal Register: January 31, 2017 (82 FR 8871).

The Commission's related evaluation of the amendment and final no significant hazards consideration determination are contained in a Safety Evaluation dated March 9, 2017.

No significant hazards consideration comments received: No.

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

Date amendment request: May 17, 2016, as supplemented by letters dated November 2, 2016, and March 1, 2017.

Brief description of amendment: The amendment revised and removed certain requirements from the Section 6, “Administrative Controls,” portions of the Oyster Creek Nuclear Generating Station Technical Specifications (TSs) that are not applicable to the facility in a permanently defueled condition. In addition, the amendment added definitions to TS Section 1, “Definitions.” Also, the amendment made additions to, deletions from, and conforming administrative changes to the TSs.

Date of issuance: March 7, 2017.

Effective date: Effective upon the licensee's submittal of the certifications required by 10 CFR 50.82(a)(1)(i) and 50.82(a)(1)(ii), and shall be implemented within 60 days of the effective date of the amendment, but may not exceed March 29, 2020.

Amendment No.: 290. A publicly-available version is in ADAMS under Accession No. ML16235A413; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-16: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: July 19, 2016 (81 FR 46963). On July 19, 2016, the NRC staff published a proposed no significant hazards consideration (NSHC) determination regarding the amendment request in the Federal Register (81 FR 46963). Subsequently, by letter dated November 2, 2016, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the NRC staff published a second proposed NSHC determination regarding the amendment request in the Federal Register on November 22, 2016 (81 FR 83876), which superseded the original Federal Register notice in its entirety. The supplemental letter dated March 1, 2017, provided additional information that clarified the application, did not expand the scope of the application as noticed, and did not change the NRC staff's second proposed NSHC determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 7, 2017.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 50-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia

Date of amendment request: June 16, 2016.

Brief description of amendments: The amendments changed Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric Generating Plant Units 3 and 4. The amendments authorized 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. Specifically, the changes to the Technical Specifications (TS) and information in the UFSAR revised the AP1000 protection and safety monitoring system functional logic to comply with the requirements on operating bypasses in Clause 6.6, “Operating Bypasses” of the Institute of Electrical and Electronics Engineers (IEEE) Std. 603-1991, “IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations.”

Date of issuance: February 24, 2017.

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

Amendment Nos.: 71/70. A publicly-available version is in ADAMS under Accession No. ML16320A097; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined License and TS.

Date of initial notice in Federal Register: August 16, 2016 (81 FR 54610).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 24, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 30, 2016, as supplemented by letter dated August 4, 2016.

Brief description of amendment: This amendment revised the date of the Cyber Security Plan implementation schedule for Milestone 8. Milestone 8 requires full implementation of the VCSNS Cyber Security Plan.

Date of issuance: March 9, 2017.

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

Amendment No.: 208. A publicly-available version is in ADAMS under Accession No. ML17011A050; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-12: Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: October 4, 2016 (81 FR 68472).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 9, 2017.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 16th day of March 2017.

For the Nuclear Regulatory Commission.

Kathryn M. Brock,

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