81 Fed. Reg. 46958 (Jul. 19, 2016)

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

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Federal RegisterJul 19, 2016

81 Fed. Reg. 46958 (Jul. 19, 2016)

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from June 21, 2016, to July 1, 2016. The last biweekly notice was published on July 5, 2016 (81 FR 43646).

DATES:

Comments must be filed by August 18, 2016. A request for a hearing must be filed by September 19, 2016.

ADDRESSES:

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

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID: NRC-2016-0141. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

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

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID: NRC-2016-0141 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID: NRC-2016-0141.
NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section.
NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2016-0141, facility name, unit number(s), application date, and subject in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov,, as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

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

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, 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 person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with NRC regulations, policies and procedures.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii). If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of any amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

A State, local governmental body, federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by September 19, 2016. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions for leave to intervene set forth in this section, except that under § 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c).

If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by September 19, 2016.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least ten 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC's Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html,, by email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/ehd/,, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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

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

Date of amendment request: May 25, 2016. A publicly available version is in ADAMS under Accession No. ML16146A639.

Description of amendment request: The amendment would replace the CR-3 Permanently Defueled Emergency Plan and its associated Emergency Action Level (EAL) Bases Manual with the Independent Spent Fuel Storage Installation (ISFSI)-Only Emergency Plan (IOEP) and its associated EAL Bases Manual. This IOEP will be used at CR-3 after all spent fuel has been transferred to the CR-3 ISFSI.

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?

The proposed amendment would modify the CR-3 facility operating license by revising the emergency plan and revising the EAL scheme. CR-3 has permanently ceased operation and is permanently defueled. The proposed amendment is conditioned on all spent nuclear fuel being removed from wet storage in the spent fuel pools and placed in dry storage within the ISFSI. Occurrence of postulated accidents associated with spent fuel stored in a spent fuel pool is no longer credible in a spent fuel pool devoid of such fuel. The proposed amendment has no effect on plant systems, structures, or components (SSC) and no effect on the capability of any plant SSC to perform its design function. The proposed amendment would not increase the likelihood of the malfunction of any plant SSC. The proposed amendment would have no effect on any of the previously evaluated accidents in the CR-3 Final Safety Analysis Report.

Since CR-3 has permanently ceased operation, the generation of fission products has ceased and the remaining source term continues to decay. This continues to significantly reduce the consequences of previously evaluated postulated accidents. Therefore, the proposed amendment does not involve a significant increase in the consequences of a previously evaluated accident.

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

The proposed amendment constitutes a revision of the emergency planning function commensurate with the ongoing and anticipated reduction in radiological source term at CR-3.

The proposed amendment does not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment as a result of the proposed amendment. Similarly, the proposed amendment would not physically change any SSC involved in the mitigation of any postulated accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed amendment does not create the possibility of a new failure mode associated with any equipment or personnel failures. The credible events for the ISFSI remain unchanged.

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?

Because the 10 CFR part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. With all spent nuclear fuel transferred out of wet storage from the spent fuel pools and placed in dry storage within the ISFSI, a fuel handling accident is no longer credible. There are no longer credible events that would result in radiological releases beyond the site boundary exceeding the EPA [Environmental Protection Agency] Protective Action Guide exposure levels, as detailed in the EPA's “Protective Action Guide and Planning Guidance for Radiological Incidents,” Draft for Interim Use and Public Comment dated March 2013 (PAG [Protective Action Guide] Manual).

The proposed amendment does not involve a change in the plant's design, configuration, or operation. The proposed amendment does not affect either the way in which the plant structures, systems, and components perform their safety function or their design margins. Because there is no change to the physical design of the plant, there is no change to these margins.

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: Lara S. Nichols, 550 South Tryon Street, Charlotte, NC 28202.

NRC Branch Chief: Bruce A. Watson.

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

Date of amendment request: May 10, 2016, as supplemented by letter dated May 18, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16131A891 and ML16139A161, respectively.

Description of amendment request: The amendment would revise the safety function lift and lower setpoint tolerances of the safety/relief valves (SRVs) that are listed in Surveillance Requirements 3.4.3.1 and 3.4.4.1 of the Technical Specifications (TSs).

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

Response: No.

This proposed amendment has no influence on the probability or consequences of any accident previously evaluated. The lower safety setpoint tolerance change does not affect the operation of the SRVs and it does not affect the as-left setpoint tolerance band which is unchanged at ±3% of the lift setpoint of the SRVs. The change only affects the lower tolerance for opening of the SRVs. The proposed amendment does not affect the upper tolerance for SRVs safety setpoints, which is the limit that protects from overpressurization.

The proposed amendment does not involve any physical changes to the SRVs, nor does it change the safety function of the SRVs. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions as discussed in the technical evaluation for this [license amendment request]. Additionally, the proposed change does not involve any significant changes to existing structures, systems, or components.

The proposed amendment does not change any other behavior or operation of the SRVs, and, therefore, has no significant impact on reactor operation. It also has no significant impact on response to any perturbation of reactor operation including transients and accidents previously analyzed in the [Final Safety Analysis Report (FSAR)].

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

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

Response: No.

The proposed change from −3% to −5% for the SRV safety setpoint lower tolerance only affects the criteria to determine when an as-found SRV test is considered acceptable. The proposed change does not affect the criteria for the setpoint upper tolerance for the SRVs.

The proposed change from −3% to −5% for the SRV safety setpoint lower tolerance does not adversely affect the operation of any safety-related components or equipment. Since the proposed amendment does not involve any hardware changes, significant changes to the operation of any systems or components, nor change to existing structures, systems, or components, there is no possibility that a new or different kind of accident is created.

The proposed change from −3% to −5% for the SRV safety setpoint lower tolerance does not involve any physical changes to the SRVs, nor does it change the safety function of the SRVs. The proposed change does not require any physical change or alteration of any existing plant equipment. No new or different equipment is being installed. No installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated. This change does not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed. No changes are being made to the procedures relied upon to respond to off-normal events as described in the FSAR are being proposed by this change. The proposed change does not alter assumptions made in the safety analysis and licensing basis.

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

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

Response: No.

The proposed change from −3% to −5% for the SRV safety setpoint lower tolerance only affects the criteria to determine when an as-found SRV test is considered acceptable. This change does not affect the criteria for the SRV safety setpoint upper tolerance. The TS setpoints for the SRVs are not changed. The as-left setpoint tolerances are not changed by the proposed amendment and remain at ±3%.

The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change from −3% to −5% for the SRV safety setpoint lower tolerance does not significantly impact the condition or performance of structures, systems, and components relied upon for accident mitigation.

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

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

NRC Branch Chief: Robert J. Pascarelli.

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: June 20, 2016. A publicly available version is in ADAMS under Accession No. ML16173A371.

Description of amendment request: The amendments would revise the Technical Specification (TS) requirements associated with the storage inventory of lube oil for the emergency diesel generators (EDGs). Specifically, the TS volume requirements for stored EDG lube oil (currently specified in number of gallons) would be replaced with volume requirements based on EDG operating time (specified in number of days). The volume requirements, specified in number of gallons, along with the equivalent number of days of EDG operating time, would be included in the TS Bases. As such, the amendments would allow the licensee to make changes to the number of gallons using the provisions of 10 CFR 50.59, consistent with the TS Bases Control Program specified in TS 5.5.10. The proposed changes are based on Revision 1 to Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-501, “Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control.”

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

Response: No.

The proposed change relocates the volume of diesel lube oil required to support 7-day operation of each onsite diesel generator, and the volume equivalent to a 6-day supply, to licensee control. The specific volume of lube oil equivalent to a 7-day and 6-day supply is based on the diesel generator manufacturer's consumption values for the run time of the diesel generator. Because the requirement to maintain a 7-day supply of diesel lube oil is not changed and is consistent with the assumptions in the accident analyses, and the actions taken when the volume of lube oil is less than a 6-day supply have not changed, neither the probability nor the consequences of any accident previously evaluated will be affected.

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

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

Response: No.

The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change does not alter assumptions made in the safety analysis but ensures that each diesel generator operates as assumed in the accident analysis. The proposed change is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

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

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

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

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: May 17, 2016. A publicly-available version is in ADAMS under Accession No. ML16138A129.

Description of amendment request: The proposed amendment would revise OCNGS's Technical Specification (TS) Section 6.0, “Administrative Controls.”

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

Response: No.

The proposed changes would not take effect until OCNGS has permanently ceased operation and entered a permanently defueled condition. The proposed changes would revise the OCNGS TS by deleting or modifying certain portions of the TS administrative controls described in Section 6.0 of the TS that are no longer applicable to a permanently shutdown and defueled facility.

The proposed changes do not involve any physical changes to plant Structures, Systems, and Components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting control settings, limiting conditions for operation, surveillance requirements, or design features.

The deletion and modification of provisions of the administrative controls do not directly affect the design of SSCs necessary for safe storage of spent irradiated fuel or the methods used for handling and storage of such fuel in the Spent Fuel Pool (SFP). The proposed changes are administrative in nature and do not affect any accidents applicable to the safe management of spent irradiated fuel or the permanently shutdown and defueled condition of the reactor.

In a permanently defueled condition, the only credible accidents are the Fuel Handling Accident (FHA), Radioactive Liquid Waste System Leak, and Postulated Radioactive Releases Due to Liquid Tank Failures. Other accidents such as Loss of Coolant Accident, Loss of Feedwater, and Reactivity and Power Distribution Anomalies will no longer be applicable to a permanently defueled reactor plant.

The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a permanently defueled condition will be the only operation allowed, and therefore, 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 changes do not involve a significant increase in the probability or consequence of an accident previously evaluated.

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

Response: No.

The proposed changes to delete and/or modify certain TS administrative controls have no impact on facility SSCs affecting the safe storage of spent irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of spent irradiated fuel itself. The proposed changes do not result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shut down and defueled and OCNGS will no longer be authorized to operate the reactor.

The proposed changes do not affect systems credited in the accident analysis for the FHA, Radioactive Liquid Waste System Leak, and Postulated Radioactive Releases Due to Liquid Tank Failures at OCNGS. The proposed changes will continue to require proper control and monitoring of safety significant parameters and activities. The proposed changes do not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers in support of maintaining the plant in a permanently shutdown and defueled condition (e.g., fuel cladding and SFP cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

The proposed changes do not alter the protection system design, create new failure modes, or change any modes of operation. The proposed changes do not involve a physical alteration of the plant, and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident.

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 involve deleting and/or modifying certain TS administrative controls once the OCNGS facility has been permanently shutdown and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for OCNGS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel following submittal of the certifications required by 10 CFR 50.82(a)(1). As a result, the occurrence of certain design basis postulated accidents are no longer considered credible when the reactor is permanently defueled.

The only remaining credible accident is a fuel handling accident (FHA). The proposed changes do not adversely affect the inputs or assumptions of any of the design basis analyses that impact the FHA.

The proposed changes are limited to those portions of the TS administrative controls that are related to the safe storage and maintenance of spent irradiated fuel. The requirements that are proposed to be revised and/or deleted from the OCNGS TS are not credited in the existing accident analysis for the remaining applicable postulated accident (i.e., FHA); therefore, they do not contribute to the margin of safety associated with the accident analysis. Certain postulated DBAs [design-basis accidents] involving the reactor are no longer possible because the reactor will be permanently shut down and defueled and OCNGS will no longer be authorized to operate the reactor.

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

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

NRC Acting Branch Chief: Shaun M. Anderson.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, Texas

Date of amendment request: April 27, 2016. A publicly available version is in ADAMS under Accession No. ML16120A432.

Description of amendment request: The amendments would revise the Technical Specifications (TSs) by eliminating Section 5.5.8, “Inservice Testing Program,” and adding a new defined term, “Inservice Testing Program,” to the TS Definitions section. The proposed amendments are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing,” dated October 21, 2015 (ADAMS Accession No. ML15294A555).

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

Response: No.

The proposed change revises TS Chapter 5, “Administrative Controls,” Section 5.5, “Programs and Manuals,” by eliminating the “Inservice Testing Program” specification. Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the [American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code], as clarified by Code Case OMN-20, “Inservice Test Frequency.” The remaining requirements in the Section 5.5.8 [Inservice Testing (IST)] Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, “Inservice Testing Program,” is added to the TS, which references the requirements of 10 CFR 50.55a(f).

Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.

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

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

Response: No.

The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged. However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

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

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

Response: No.

The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate existing TS SR 3.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety.

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

Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.

NRC Branch Chief: Robert J. Pascarelli.

NextEra Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: March 31, 2016, as supplemented by letter dated May 31, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16095A278 and ML16159A194, respectively.

Description of amendment request: The amendment would revise Technical Specification (TS) 6.15, “Containment Leakage Rate Testing Program,” to require a program that is in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J” (ADAMS Accession No. ML12221A202). The proposed change would allow extension of the Type A test interval up to one test in 15 years, and extension of the Type C test interval up to 75 months, based on acceptable performance history as defined in NEI 94-01, Revision 3-A.

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

Response: No.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, “Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,” for development of the Seabrook performance-based containment testing program. NEI 94-01 allows, based on risk and performance, an extension of Type A and Type C containment leak test intervals. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses.

The findings of the Seabrook risk assessment confirm the general findings of previous studies that the risk impact with extending the containment leak rate is small. Per the guidance provided in Regulatory Guide 1.174, an extension of the leak test interval in accordance with NEI 94-01, Revision 3-A results in an estimated change within the small change region.

Since the change is implementing a performance-based containment testing program, the proposed amendment does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The requirement for containment leakage rate acceptance will not be changed by this amendment. Therefore, the containment will continue to perform its design function as a barrier to fission product releases.

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 previously evaluated?

Response: No.

The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test frequency, does not change the design or operation of structures, systems, or components of the plant.

The proposed changes would continue to ensure containment integrity and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by these changes. Therefore, the proposed changes will not create the possibility of 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 kind of accident from any accident previously evaluated.

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

Response: No.

Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test frequency, does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation.

The specific requirements and conditions of the TS Containment Leakage Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained. This ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by implementation of a performance-based containment testing program.

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

Attorney for licensee: William Blair, Managing Attorney—Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: May 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16132A374.

Description of amendment request: The amendment would revise Technical Specification (TS) requirements by deleting TS Action Statement 3.4.2.1.b concerning stuck open safety/relief valves.

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

Response: No.

The proposed TS change deletes Action Statement 3.4.2.1.b concerning safety/relief valves. The two (2) minute action represents detailed methods of responding to an event, and therefore, if eliminated, would not result in increasing the probability of the event, nor act as an initiator of an event. Limiting condition for operation 3.6.2.1, “Depressurization Systems—Suppression Chamber,” and plant procedures provide operators with appropriate direction for response to a suppression pool high temperature (which could be caused by a stuck open relief valve). Providing specific direction to close the valve within two (2) minutes does not provide additional plant protection beyond what is provided for in plant procedures and TS 3.6.2.1.

Therefore, this action can be eliminated, and will not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed TS change deletes Action Statement 3.4.2.1.b concerning safety/relief valves. This change does not change the design or configuration of the plant. No new operation or failure modes are created, nor is a system-level failure mode created that is different than those that already exist.

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

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

Response: No.

The proposed change does not involve a significant reduction in a margin of safety, nor does it affect any analytical limits. There are no changes to accident or transient core thermal hydraulic conditions, or fuel or reactor coolant boundary design limits, as a result of the proposed change. The proposed change will not alter the assumptions or results of the analysis contained in the Updated Final Safety Analysis Report (UFSAR).

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

Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: Douglas A. Broaddus.

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: May 17, 2016. A publicly-available version is in ADAMS under Accession No. ML16138A431.

Description of amendment request: The amendment request proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document (DCD) Tier 2 information and involves changes to related Tier 1 information, with corresponding changes to the associated Combined License (COL) Appendix C 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, “Design Certification Rule for the AP1000 Design,” is also requested for the plant-specific DCD Tier 1 material departures. Specifically, the requested amendment proposes changes to the concrete wall thickness tolerance for the column line N wall, from column lines 2 to 4 from elevation 100′-0″ to 135′-3″, from plus or minus 1 inch to plus 4 inches.

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

Response: No.

As indicated in the UFSAR Subsection 3.8.4.1.2, the auxiliary building contains structural modules in the south side of the building that include the spent fuel pool, fuel transfer canal, and cask loading and washdown pits. The increase in tolerance associated with the concrete thickness of the concrete wall for the column line N from column line 2 to 4 and the deviation from ACI 349-01 does not involve any accident initiating components or events, thus leaving the probabilities of an accident unaltered. The increased tolerance does not adversely affect any safety-related structures or equipment nor does the increased tolerance reduce the effectiveness of a radioactive material barrier. Thus, the proposed changes would not affect any safety-related accident mitigating function served by the containment internal structures.

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 tolerance increase and code deviation from ACI 349-01 does not change the performance of the affected radiologically controlled portion of the auxiliary building. As demonstrated by the continued conformance to the other applicable codes and standards governing the design of the structures, and in conjunction with the analysis of a special system of construction in accordance with ACI 349-01 Section 1.4, the wall with an increased concrete thickness tolerance continues to withstand the same effects as previously evaluated. There is no change to the design function of the affected module and wall, and no new failure mechanisms are identified as the same types of accidents are presented to the wall before and after the 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.

The proposed change to increase the concrete thickness tolerance for the column line N wall from column line 2 to 4 identified in COL Appendix C Table 3.3-1 does not alter any design function, design analysis, or safety analysis input or result, and sufficient margin exists to justify departure from the ACI 349-01 requirements for the wall. As such, because the system continues to respond to design basis accidents in the same manner as before without any changes to the expected response of the structure, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes. Accordingly, no safety margin is reduced by the increase of the wall concrete thickness tolerance.

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

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

NRC Acting 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: May 5, 2016. A publicly-available version is in ADAMS under Accession No. ML16126A276.

Description of amendment request: The proposed changes would revise the Combined Licenses (COLs) concerning the design details of the safety-related passive core cooling system (PXS), the nonsafety-related normal residual heat removal system (RNS), and the nonsafety-related containment air filtration system (VFS). The amendment request proposes changes to the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the plant-specific Design Control Document (DCD) Tier 2 information and involves changes to related plant-specific DCD Tier 1 information, with corresponding changes to the associated COL Appendix C information. Because this proposed change would require a departure from Tier 1 information in the Westinghouse Advanced Passive 1000 DCD, the licensee also requests an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).

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 affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSCs) accident initiator or initiating sequence of events. The proposed changes result from identifying PSX, RNS, and VFS piping lines required to be described in the licensing basis as ASME [American Society of Mechanical Engineers] Code Section III, evaluated to meet the LBB [leak-before-break] design criteria, or designed to withstand combined normal and seismic design basis loads without a loss of functional capability. Neither planned or inadvertent operation nor failure of the PXS, RNS, or VFS is an accident initiator or part of an initiating sequence of events for an accident previously evaluated. Therefore, the probabilities of the accidents evaluated in the UFSAR are not affected.

The proposed changes do not have an adverse impact on the ability of the PXS, RNS, or VFS to perform their design functions. The design of the PXS, RNS, and VFS continues to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the changes ensure that the capabilities of the PXS, RNS, and VFS to mitigate the consequences of an accident meet the applicable regulatory acceptance criteria, and there is no adverse effect on any safety-related SSC or function used to mitigate an accident. The changes do not affect the prevention and mitigation of other abnormal events, e.g., anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR are not affected.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes result from identifying PXS, RNS, and VFS piping lines required to be described in the licensing basis as ASME Code Section III, evaluated to meet the LBB design criteria, or designed to withstand combined normal and seismic design basis loads without a loss of functional capability. These proposed changes do not adversely affect any other PXS, RNS, VFS, or SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that results in significant fuel cladding failures.

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

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

Response: No.

The proposed changes maintain existing safety margins. The proposed changes ensure that PXS, RNS, and VFS design requirements and design functions are met. The proposed changes maintain existing safety margin through continued application of the existing requirements of the UFSAR, while adding additional design features to ensure the PXS, RNS, and VFS perform the design functions required to meet the existing safety margins. Therefore, the proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not adversely affect any design code, function, design analysis, safety analysis input or result, or design/safety margin. Because no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, no margin of safety is reduced.

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

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

NRC Acting Branch Chief: Jennifer Dixon-Herrity.

STP Nuclear Operating Company (STPNOC), Docket No. 50-498, South Texas Project (STP), Unit 1, Matagorda County, Texas

Date of amendment request: April 7, 2016, as supplemented by letter dated May 25, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16110A297 and ML16162A196, respectively.

Description of amendment request: The amendment would revise Technical Specification 5.3.2 for STP, Unit 1, to allow permanent operation with 56 full-length control rods with no control rod assembly in core location D-6.

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

Response: No.

STPNOC has performed a multi-cycle assessment on previous Unit 1 reactor cores and evaluated the consequences associated with removal of Control Rod D-6. The assessment indicates that removal of Control Rod D-6 does impact reactivity parameters (e.g., shutdown margin and trip reactivity); however, sufficient margin exists to ensure the Updated Final Safety Analysis Report (UFSAR) accident analysis limits continue to be met. The physical changes associated with the removal of Control Rod D-6 do not impact the probability of occurrence of a previously evaluated accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

Operation of STP Unit 1 with Control Rod D-6 removed will not create the possibility of a new or different kind of accident from any accident previously evaluated. To preserve the reactor coolant system flow characteristics in the reactor core, a flow restrictor will be installed at the top of the D-6 guide tube housing. Installation of this component will not prevent the remaining 56 control rods from performing the required design function of providing adequate shutdown margin. No new operator actions are created as a result of the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

Operation of STP Unit 1 with Control Rod D-6 removed will not involve a significant reduction in a margin of safety. The margin of safety is established by setting safety limits and operating within those limits. The proposed change does not alter a UFSAR design basis or safety limit and does not change any setpoint at which automatic actuations are initiated. STPNOC will continue to confirm all safety analysis limits remain bounding on a cycle-specific basis using an NRC-approved Westinghouse core reload evaluation methodology. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Kym Harshaw, General Counsel, STP Nuclear Operating Company, P.O. Box 289, Wadsworth, TX 77483.

NRC Branch Chief: Robert J. Pascarelli.

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

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

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

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

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

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

Date of amendment request: December 17, 2012, as supplemented by letters dated February 25, 2013; May 28, 2013; July 21, 2015; December 18, 2015; and June 1, 2016.

Brief description of amendment: The amendment revised the MPS2 Technical Specifications (TSs) to reflect the results and constraints of a new criticality safety analysis for fuel assembly storage in the MPS2 fuel storage racks.

Date of issuance: June 23, 2016.

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

Amendment No.: 327. A publicly-available version is in ADAMS under Accession No. ML16003A008; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: June 11, 2013 (78 FR 35060). The supplemental letters dated May 28, 2013; July 21, 2015; December 18, 2015; and June 1, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

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

No significant hazards consideration comments received: No.

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

Date of amendment request: August 20, 2015.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) to allow the use of Optimized Zirlo^TM. Specifically, the proposed changes modify TS 4.2.1 to add Optimized Zirlo^TM as an allowable cladding and TS 5.6.5.b to add associated methodologies for determining the core operating limits report.

Date of issuance: June 21, 2016.

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

Amendment Nos.: McGuire—288 (Unit 1) and 267 (Unit 2); Catawba—284 (Unit 1) and 280 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16105A326; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: November 24, 2015 (80 FR 73236).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 21, 2016.

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 amendment request: July 9, 2015, as supplemented by letter dated January 7, 2016.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.3.1, “Reactor Trip System (RTS) Instrumentation,” to resolve an operable but degraded non-conforming issue associated with the reactor coolant pump under-frequency trip setpoint allowable value for the McGuire Nuclear Station, Units 1 and 2.

Date of issuance: June 21, 2016.

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

Amendment Nos.: 287 (Unit 1) and 266 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16109A084; 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 Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: October 13, 2015 (80 FR 61479). The supplemental letter dated January 7, 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 June 21, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 29, 2015, as supplemented by letter dated December 3, 2015.

Brief description of amendment: The amendment revised the full implementation date (Milestone 8) of the RBS Cyber Security Plan and revised the associated license condition for the Facility Operating License. The license was also revised, in part, to include administrative and editorial corrections.

Date of issuance: June 21, 2016.

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

Amendment No.: 190. A publicly-available version is in ADAMS under Accession No. ML16124A688; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-47: The amendment revised the Facility Operating License.

Date of initial notice in Federal Register: April 5, 2016 (81 FR 19647).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 21, 2016.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana

Date of amendment request: November 17, 2011, as supplemented by letters dated January 26, September 27 and October 16, 2012; May 16, June 26, and December 18, 2013; June 11, 2014; March 12, April 10, May 14, August 27, September 8, September 24, and October 13, 2015; and January 18, 2016.

Brief description of amendment: The amendment permits the licensee to adopt a new risk-informed, performance-based fire protection licensing basis for Waterford 3, in accordance with the requirements in 10 CFR 50.48(a) and (c) and the guidance in NRC Regulatory Guide 1.205, “Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,” December 2009; National Fire Protection Association (NFPA) 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants” (2001 Edition); and Nuclear Energy Institute 04-02, “Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c),” Revision 2.

Date of issuance: June 27, 2016.

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

Amendment No.: 248. A publicly-available version is in ADAMS under Accession No. ML16126A033; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-38: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: April 10, 2012 (77 FR 21597). The supplements dated September 27 and October 16, 2012; May 16, June 26, and December 18, 2013; June 11, 2014; March 12, April 10, May 14, August 27, September 8, September 24, and October 13, 2015; and January 18, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 27, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 4, 2015, as supplemented by letters dated February 3, 2016; March 29, 2016; and June 16, 2016.

Brief description of amendment: The amendment relocated specific technical specification surveillance frequencies to a licensee-controlled program with the adoption of Technical Specification Task Force (TSTF) Traveler TSTF-425, Revision 3, “Relocate Surveillance Frequencies to Licensee Control—Risk Informed Technical Specification Task Force Initiative 5b”. Additionally, the change added a new program, the Surveillance Frequency Control Program, to Technical Specification Section 5, Administrative Controls.

Date of issuance: June 28, 2016.

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

Amendment No.: 122. A publicly-available version is in ADAMS under Accession No. ML16125A485; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-18: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: October 13, 2015 (80 FR 61482). The supplemental letters dated February 3, 2016; March 29, 2016; and 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 amendment is contained in a Safety Evaluation dated June 28, 2016.

No significant hazards consideration comments received: No.

Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida

Date of amendment request: October 6, 2015, as supplemented by letter dated March 25, 2016.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) related to moderator temperature coefficient requirements.

Date of issuance: June 20, 2016.

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

Amendment Nos: 271 (Unit No. 3) and 266 (Unit No. 4). A publicly-available version is in ADAMS under Accession No. ML16120A473; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses and TSs.

Date of initial notice in Federal Register: March 8, 2016 (81 FR 12141). The supplemental letter dated March 25, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

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

No significant hazards consideration comments received: No.

Northern States Power Company—Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: July 15, 2015.

Brief description of amendment: The amendment adopts the NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”

Date of issuance: June 21, 2016.

Effective date: As of the date of issuance and shall be implemented prior to the startup from the 2017 refueling outage.

Amendment No.: 189. A publicly-available version is in ADAMS under Accession No. ML16125A165; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-22: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: October 13, 2015 (80 FR 61484).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 21, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: May 10, 2016.

Brief description of amendments: The amendments extend the implementation period for the Salem, Unit No. 1, License Amendment No. 311, and the Salem, Unit No. 2, License Amendment No. 292, which were effective as of the date of issuance (i.e., March 7, 2016). Specifically, the implementation period for the above amendments has been extended from July 5, 2016 (i.e., 120 days from the date of issuance), to prior to entry into Mode 6 for the Salem, Unit No. 1, Fall 2017 refueling outage (1R25), and prior to entry into Mode 6 for the Salem, Unit No. 2, Spring 2017 refueling outage (2R22), to align with the outages for which the replacement of the source range and intermediate range detectors is scheduled.

Date of issuance: June 29, 2016.

Effective date: As of the date of issuance and shall be implemented by July 5, 2016.

Amendment Nos.: 314 (Unit No. 1) and 295 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML16137A579; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Renewed Facility Operating Licenses.

Date of initial notice in Federal Register: May 23, 2016 (81 FR 32351).

The Commission's related evaluation of the amendments and final no significant hazards consideration determination are contained in a Safety Evaluation dated June 29, 2016.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 4, 2016.

Brief description of amendment: The amendment revised the date of the Cyber Security Plan implementation schedule Milestone 8 and paragraph 2.E in the Facility Operating License.

Date of issuance: June 23, 2016.

Effective date: As of the date of issuance and shall be implemented within 14 days of issuance.

Amendment No.: 106. A publicly-available version is in ADAMS under Accession No. ML16146A745; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-90: Amendment revised the Facility Operating License.

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

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

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 8th day of July 2016.

For the Nuclear Regulatory Commission.

Anne T. Boland,

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

[FR Doc. 2016-16925 Filed 7-18-16; 8:45 am]

BILLING CODE 7590-01-P