NRC-2016-0180, Biweekly FRN - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations- Publication Date: August 30, 2016

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Biweekly FRN - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations- Publication Date: August 30, 2016
ML16229A255
Person / Time
Site: Millstone, Peach Bottom, Palisades, Wolf Creek, Grand Gulf, Braidwood, Summer, Vogtle, Farley, Cook, La Crosse  Entergy icon.png
Issue date: 08/18/2016
From: Boland A T
Division of Operating Reactor Licensing
To:
Rohrer S, NRR-DORL 415-5411
References
NRC-2016-0180
Download: ML16229A255 (38)
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Text

[7590-01-P]

NUCLEAR REGULATORY COMMISSION [NRC-2016-0180] 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 August 2, 2016, to August 15, 2016. The last biweekly notice was published on August 16, 2016.

DATES: Comments must be filed by [INSERT DATE 30 DAYS AFTER DATE OF PUBLICATION IN THE FEDERAL REGISTER

]. A request for a hearing must be filed by [INSERT DATE 60 DAYS AFTER DATE OF PUBLICATION IN THE FEDERAL REGISTER

].

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):

Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.

  • Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. For additional direction on obtaining information and submitting comments, see "Obtaining Information and Submitting Comments" in the SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone:

301-415-5411, e-mail:

Shirley.Rohrer@nrc.gov. I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0180, 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:

  • 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 e-mail to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.

  • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC-2016-0180, 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 considerat ion. 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 t he 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 to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the 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 the NRC's regulations, policies, and procedures. Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(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 Co mmission 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 [INSERT DATE 60 DAYS FROM THE DATE OF PUBLICATION IN THE FEDERAL REGISTER

]. 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 10 CFR 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian Tribe, or agenc y thereof may also have the opportunity to participate under 10 CFR 2.315(c). If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail 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 det ailed in the NRC's "Guidance for Electronic Submission to the NRC," which is available on the agency's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html

. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Electronic Filing Help Desk will not be able to offer assistance in using unlisted software.

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. 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/electronic-sub-ref-mat.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electron ic 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 e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail 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 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 e-mail to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 7 p.m., Eastern Time, Monday through Friday, excluding government holidays. Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) first class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 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 proc eedings 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 hearing request and petition to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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

Dominion Nuclear Connecticut, Inc. (DNC), Docket No. 50-336, Millstone Power Station, Unit No. 2 (MPS2), New London County, Connecticut Date of amendment request: May 25, 2016. A publicly-available version is in ADAMS under Accession No. ML16153A026. Description of amendment request: The amendment would add the AREVA topical report, EMF-2103(P)(A), "Realistic Large Break [loss of coolant accident] LOCA [RLBLOCA] Methodology for Pressurized Water Reactors," Revision 3, to MPS2 Technical Specification (TS) 6.9.1.8.b, "Core Operating Limits Report," which lists the analytical methods used to determine the core operating limits. The methodology in EMF-2013(P)(A) for RLBLOCA has been used for the MPS2 LBLOCA analysis of the AREVA Standard CE-14 HTP fuel product with M5 cladding, which DNC plans to introduce beginning with the fresh fuel for MPS2 Cycle 25 in spring 2017. 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 revisions provided in [brackets]: 1. Does the proposed [amendment] involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change to TS 6.9.1.8.b permits the use of the AREVA RLBLOCA methodology to analyze the MPS2 LBLOCA to ensure that the plant continues to meet the Emergency Core Cooling System (ECCS) performance acceptance criteria in 10 CFR 50.46. The RLBLOCA analysis demonstrates MPS2 continues to satisfy the 10 CFR 50.46 ECCS performance acceptance criteria using an NRC-approved

evaluation model. The proposed change to the list of NRC-approved methodologies listed in TS 6.9.1.8.b has no impact on how the plant is operated or configured. Addition of this methodology to the list of methodologies in TS 6.9.1.8.b does not impact either the probability or consequences of an accident currently evaluated in Chapter 14 of the [Updated Final Safety Analysis Report] UFSAR.

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] cr eate the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

The proposed change to TS 6.9.1.8.b adds topical report EMF-2103(P)(A) to the list of approved methodologies for determining core operating limits at MPS2. The proposed amendment has no adverse effect on plant operation or accident mitigation equipment. The amendment does not create any new credible failure mechanisms, malfunctions, or accident initiators not considered in the current design basis accidents (DBAs). The response of the plant and operators following a DBA will not be changed. The proposed amendment does not create the possibility of a new failure mode associated with any equipment or human performance failures. Thus, the possibility of a new or different type of accident is not

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

FSAR.

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

Response: No.

The proposed change to TS 6.9.1.8.b adds topical report EMF-2103(P)(A) to the list of approved methodologies for determining core operating limits at MPS2. Approved methodologies will be used to ensure that the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed amendment has no [e]ffect on the ability of the plant to mitigate DBAs and ensure consequences of the existing DBA remains bounding. The margin of safety to mitigate consequences of DBAs is not reduced. Structures, systems and components used to mitigate DBAs are not affected. No changes are

being made to safety limits or safety system settings required by TS. Therefore, the proposed change does not involve a significant reduction

in a margin of safety.

The NRC staff has reviewed the licensee's ana lysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lillian M. Cuoco, Seni or Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219. NRC Branch Chief: Travis L. Tate.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant (PNP), Van Buren County, Michigan Date of amendment request: July 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16193A005. Description of amendment request: The propos ed amendment would revise the Technical Specifications (TSs) to eliminate TS Section 5.5.7, "Inservice Testing [IST] Program." A new defined term, "INSERVICE TESTING PROGRAM," is added to the TS Definitions section. This amendment request is consistent with Technical Specifications Task Force (TSTF)-545, Revision 3, "TS Inservice Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing." 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 change would revise TS Chapter 5, Administrative Controls, Section 5.5, Programs and Manuals, by eliminating the TS 5.5.7, Inservice Testing Program, specification. Most requirements in the IST Program would be removed, as they are duplicative of requirements in the ASME [American Society of Mechanical Engineers] OM Code

[ASME Code for Operation and Maintenance of Nuclear Power Plants], as clarified by Code Case OMN-20, Inservice Test Frequency. The remaining requirements in the Section 5.5 IST Program would be eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, INSERVICE TESTING PROGRAM, would be added to the TS, which references the

requirements of 10 CFR 50.55a(f), Performance of IST 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. IST 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 the 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 change create the possibility of a new or different kind of accident from any [acc ident] 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 IST performed. In most cases, the frequency of IST would be 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 accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety? Response: No.

The proposed change would eliminate 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 would allow 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 would eliminate the 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 would 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 would eliminate 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.

The NRC staff has reviewed the licensee's ana lysis 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: Jeanne Cho, Senior Couns el, Entergy Services, Inc., 440 Hamilton Ave., White Plains, NY 10601. NRC Branch Chief: David J. Wrona.

LaCrosseSolutions, Inc., and Dairyland Power Cooperative, Docket Nos.: 50-409 and 72-046, La Crosse Boiling Water Reactor (LACBWR), La Crosse County, Wisconsin Date of amendment request: June 27, 2016. A publicly-available version is in ADAMS under Accession No. ML16200A083.

Description of amendment request: The propos ed amendment would amend the Possession Only License for the LACBWR to reflect the approval of the LACBWR License Termination Plan (LTP) when that review and approval process is completed by the NRC staff. The LTP will become a supplement to LACBWR's other decommissioning documents and will be implemented by the licensee to complete decommissioning activities at the LACBWR site. Once decommissioning is complete, a separate request will be made to the NRC by the licensee to terminate the LACBWR license. 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?

The only remaining accident following completion of fuel transfer to the [Independent Spent Fuel Storage Installation] (ISFSI) is a radioactive release accident where spontaneous release of the (non-ISFSI-related) radioactive source term remaining at the LACBWR site in a form and quantity is immediately released through an airborne or liquid release

path. A radioactive release analysis was performed to establish the bounding event at the site considering the current stage of LACBWR decommissioning. 1.175 [Curies] (Ci) of radioactive material is conservatively estimated in the analysis to be present on plant surfaces, and as such represents the assumed total non-ISFSI radioactive source term remaining at the LACBWR site. The LACBWR analysis of postulated release events separately considers the portion of this remaining radioactive contamination that is immediately releasable as airborne contamination and that is immediately releasable as contaminated liquid.

A conservative fraction of 30 percent of the total remaining source term is assumed in the analysis to be immediately available for airborne release.

The analysis results demonstrate that the consequences of releasing 30 percent of the non-ISFSI radioactive source term remaining at the LACBWR site to the atmosphere are well within the applicable 10 CFR 100.11 and [U.S. Environmental Protection Agency] (EPA) [Protective

Action Guides] (PAG) limits.

The portion of the total remaining source term conservatively assumed in the analysis to be available for liquid release at any one time is 80 percent of the radioactively contaminated liquid stored in the site retention tank. In the unlikely event that 80 percent of the retention tank volume at a total radionuclide concentration of 3.9E-03 Ci/cc were to be released from the retention tank at a flow rate of 20 [gallons per minute] (gpm), the normal effluent concentration limits of 10 CFR 20, Appendix B, Table 2, would not be exceeded. Thus, the liquid release analysis demonstrates that there is no reasonable likelihood that a postulated radioactive liquid release event could result in exceeding the normal effluent concentration limits of 10 CFR 20, Appendix B.

With consideration for the current stage of LACBWR decommissioning and with spent nuclear fuel now stored in the ISFSI, the bounding radioactive release analysis, for both airborne and liquid releases, confirms that the minimal radioactive material resulting from LACBWR operation and remaining on the LACBWR site is insufficient for any potential event to result in exceeding dose limits or otherwise involving a significant adverse effect on public health and safety.

The proposed change does not affect the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and has no impact on plant operations. The proposed changes do not have an adverse impact on the remaining decommissioning activities or any decommissioning related postulated accident consequences.

The proposed changes related to the approval of the LTP do not affect operating procedures or administrative controls that have the function of preventing or mitigating the remaining decommissioning design basis accident.

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

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

The accident analysis for the facility related to decommissioning activities is described in the [Decommissioning Plan / Post-Shutdown Decommissioning Activities Report] (D-Plan / PSDAR). The requested license amendment is consistent with the plant activities described in the D-Plan / PSDAR. Thus, the proposed changes do not affect the remaining plant systems, structures, or components in a way not

previously evaluated.

There are sections of the LTP that refer to the decommissioning activities still remaining. These activities are performed in accordance with approved site processes and undergo a 10 CFR 50.59 review as required prior to initiation. The proposed amendment merely makes mention of these processes and does not bring about physical changes to the facility.

Therefore, the facility conditions for which the remaining postulated accident has been evaluated is still valid and no new accident scenarios, failure mechanisms, or single failures are introduced by this amendment.

The system operating procedures are not affected. Therefore, the proposed changes will 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? The LTP is a plan for demonstrating compliance with the radiological criteria for license termination as provided in 10 CFR 20.1402. The margin of safety defined in the statements of consideration for the final rule on the Radiological Criteria for License Termination is described as the margin between the 100 [millirem per year] (mrem/yr) public dose limit established in 10 CFR 20.1301 for licensed operation and the 25 mrem/yr dose limit to the average member of the critical group at a site considered acceptable for unrestricted use (one of the criteria of 10 CFR 20.1402). This margin of safety accounts for the potential effect of multiple sources of radiation exposure to the critical group. Since the License Termination Plan is designed to comply with the radiological criteria for license termination for unrestricted use, the LTP supports this margin of safety.

In addition, the LTP provides the methodologies and criteria that will be used to perform remediation activities of residual radioactivity to demonstrate compliance with the [As Low As Reasonably Achievable]

(ALARA) criterion of 10 CFR 20.1402.

Additionally, the LTP is designed with recognition that (a) the methods in MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual) and (b) the building surface contamination levels are not directly applicable to use with complex nonstructural components. Therefore, the LTP states that nonstructural components remaining in buildings (e.g., pumps, heat exchangers, etc.) will be evaluated against the criteria of Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors," to determine if the components can be released for unrestricted use. The LTP also states that materials, surveyed and evaluated as a-part of normal decommissioning activities and prior to implementation of the final radiation surveys, will be surveyed for release using current site procedures to demonstrate compliance with the "no detectable" criteria. Such materials that do not pass these criteria will be controlled as contaminated.

Also, as previously discussed, the bounding radioactive release accident analysis for decommissioning is based on a conservative estimate of the radioactive material remaining onsite. Since the bounding accident results in a release of more airborne and liquid radioactivity than can be released from planned LTP decommissioning events, the margin of safety associated with the consequences of decommissioning accidents is not reduced by this activity.

Thus, the proposed change does not involve a significant reduction in the

margin of safety. The NRC staff has reviewed the licensee's ana lysis 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: Russ Workman, General Counsel, Energy Solutions, 299 South Main Street, Suite 1700, Salt Lake City, Utah 84111. NRC Branch Chief: Bruce Watson.

South Carolina Electric and Gas Company and South Carolina Public Service Authority, Docket Nos.52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina Date of amendment request: June 16, 2016, as revised August 8, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16168A257 and ML16221A649, respectively. Description of amendment request: The requested amendment proposes to depart from approved AP1000 Design Control Document Tier 2* and associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR). Specifically, the requested amendment proposes to depart from UFSAR text and figures that describe the connections between floor modules and structural wall modules in the containment internal structures. 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 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 design details for the floor modules and the connections between floor modules and the structural wall modules, and the change to more clearly state the design requirement that these connections meet criteria and requirements of American Concrete Institute (ACI) 349 and American Institute of Steel Construction (AISC)

N690, do not have an adverse impact on the response of the nuclear island structures to safe shutdown earthquake ground motions or loads due to anticipated transients or postulated accident conditions. The change of the design details for the connections between floor modules and the structural wall modules, and the clarification of design requirements for these connections, 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 design details for the floor modules and the connections between floor modules and the structural wall modules, and more clearly state the design requirement that these connections meet criteria and requirements of ACI 349 and AISC N690. The clarification and changes to the design details for the floor modules and the connections between floor modules and the structural wall modules do not change the design requirements of the nuclear island structures. The clarification and changes of the design details for the floor modules and the connections between floor modules and the structural wall modules 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 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. The acceptance limits for the design of seismic Category I structures are included in the codes and standards used for the design, analysis, and construction of the structures. The two primary codes for the seismic Category I structures are American Institute of Steel Construction (AISC) N690 and American Concrete Institute (ACI) 349.

The changes to the design of the connection of the floor module to the structural wall modules in the containment internal structures satisfy applicable provisions of AISC N690 and ACI 349 and supplemental requirements included in the UFSAR.

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

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

Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue, NW, Washington, DC, 20004-2514. Acting NRC Branch Chief: Jennifer Dixon-Herrity.

South Carolina Electric and Gas Company and South Carolina Public Service Authority, Docket Nos.52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina Date of amendment request: July 19, 2016. A publicly-available version is in ADAMS under Accession No. ML16202A035. Description of amendment request: The amendment request proposes changes to the Technical Specifications and Updated Final Safety Analysis Report (UFSAR) Tier 2 information to update the Protection and Safety Monitoring System (PMS) to align with the requirements in Institute of Electrical and Electronics Engineers (IEEE) 603-1991, "IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations." IEEE 603-1991, Clause 6.6, "Operating Bypasses," imposes requirements on the operating bypasses (i.e., "blocks" and "resets") used for the AP1000 PMS. The PMS functional logic for blocking the source range neutron flux doubling signal shown in UFSAR Figure 7.2-1 (Sheet 3) requires revision to fully comply with this requirement.

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 change modifies the PMS logic used to terminate an inadvertent boron dilution accident which results in a source range flux doubling signal. An inadvertent boron dilution is caused by the failure of the demineralized water transfer and storage system or chemical and volume control system, either by controller, operator or mechanical failure. The proposed changes to PMS and Technical Specification requirements do not adversely affect any of these accident initiators or introduce any component failures that could lead to a boron dilution event; thus the probabilities of accidents previously evaluated are not affected. The proposed changes do not adversely interface with or

adversely affect any system contai ning 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 accident analysis evaluates events involving a decrease in reactor coolant system boron concentration due to a malfunction of the chemical and volume control system in Modes 1 through 6. The Technical Specifications currently provide administrative controls to prevent a boron dilution event in Mode 6. The proposed change would provide additional PMS interlocks and administrative controls for prevention of a boron dilution event applicable in Modes 2, 3, 4, and 5. The proposed changes to the PMS design do not adversely affect the design or operation of safety related equipment or equipment whose failure could initiate an accident from what is already described in the licensing basis. These changes do not adversely affect fission product barriers. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change.

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

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

Response: No.

The proposed change would add additional restrictions on the source range flux doubling signal operational bypass to align it with the requirements in IEEE 603 and provide assurance that the protection logic is enabled whenever the plant is in a condition where protection might be required. These changes to the PMS design do not adversely impact nor affect the design, construction, or operation of any plant [structure, system, and components (SSCs)], including any equipment whose failure could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected by the proposed changes. Furthermore, no system function, design function, or equipment qualification will be adversely affected by the changes.

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

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

Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue, NW, Washington, DC, 20004-2514. Acting 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: July 25, 2016. A publicly-available version is in ADAMS under Accession No. ML16207A340. Description of amendment request: The amendment request proposes changes to a plant-specific Tier 1 (and combined license Appendix C) table and the Updated Final Safety Analysis Report (UFSAR) tables to clarify the flow area for the Automatic Depressurization System (ADS) fourth stage squib valves and to reduce the minimum effective flow area for the second and third stage ADS control valves. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR Part 52, Appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures.

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 do not adversely affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) accident initiator or initiating sequence of events. The proposed changes do not adversely affect the physical design and operation of the second and third stage ADS control valves and fourth stage ADS squib valves, including as-installed inspections, testing, and maintenance requirements, as described in the UFSAR. Therefore, the operation of the second and third stage ADS control valves and fourth stage ADS squib valves is not adversely

affected.

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

affected.

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

evaluated.

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

Response: No.

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

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

3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No.

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

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

The NRC staff has reviewed the licensee's ana lysis 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 Acting Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Inc., Docket No. 50-425, Vogtle Electric Generating Plant, Unit 2, Burke County, Georgia Date of amendment request: August 12, 2016. A publicly-available version is in ADAMS under Accession No. ML16225A619. Description of amendment request: The licensee proposes to modify the Vogtle Electric Generating Plant, Unit 2, Technical Specifications (TSs) Limiting Condition for Operation 3.7.9, "Ultimate Heat Sink (UHS)," such that with the 2B Nuclear Service Cooling Water (NSCW) transfer pump inoperable for refurbishment, the Completion Time of Condition 3.7.9.D.2.2 would be 46 days as opposed to 31 days. This TS change would be a one-time change and in effect only for the 2B NSCW transfer pump for the remainder of Cycle 19. 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 change does not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. The proposed changes will not alter assumptions relative to the mitigation of an accident or transient event. Furthermore, the UHS will remain capable of adequately responding to a design basis event during the period of the extended CT [Completion Time]. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not introduce any new or unanalyzed modes of operation. The refurbishment of the pump does not involve any unanalyzed modifications to the design or operational limits of the NSCW system. The redundant pump and compens atory measures allowed by the Technical Specifications will remain unaffected. Therefore, no new failure modes or accident precursors are created due to the pump refurbishment during the extended Completion Time. For the reasons noted above, the proposed change will not create the possibility of a new or different accident previously evaluated.

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

Response: No.

The margin of safety is related to the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment. The performance of these fission product barriers will not be affected by the proposed change; therefore, the margin to the onsite and offsite radiological dose limits are not significantly reduced.

During the extended Completion Time for the 2B NSCW transfer pump, the NSCW system and the UHS will remain capable of mitigating the consequences of a design basis event such as a LOCA [loss-of-coolant accident]. Technical Specifications Action 3.7.9.D.2.1 will be taken to provide an alternate method of basin transfer.

For the reasons noted above, there is no significant reduction in a margin

of safety.

The NRC staff has reviewed the licensee's ana lysis 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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

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. 50-423, Millstone Power Station, Unit No. 3 (MPS3), New London County, Connecticut Date of amendment request: August 31, 2015. Brief description of amendment: The amendment revised the MPS3 Design Features - Fuel Storage Technical Specification 5.6.3, "Capacity," to specify the spent fuel pool storage capacity limit in terms of the total number of fuel assemblies. Date of issuance: August 4, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment No.: 270. A publicly-available version is in ADAMS under Accession No. ML16206A001; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. NPF-49: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register

November 24, 2015 (80 FR 73235). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 4, 2016. No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi Date of application for amendment: September 15, 2015. Brief description of amendment: The amendment revised the GGNS Technical Specifications (TSs) to eliminate the "Inservice Testing Program," specification in Section 5.5, "Programs and Manuals," which is superseded by Code Case OMN-20. A new defined term, "INSERVICE TESTING PROGRAM," would be added to TS Section 1.1, "Definitions." This request is consistent with TS Task Force (TSTF)-545, Revision 1, "TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing."

Date of issuance: August 4, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment No: 211. A publicly-available version is in ADAMS under Accession No. ML16140A133; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. NPF-29: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: March 1, 2016 (81 FR 10679). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 4, 2016. No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station (Braidwood), Units 1 and 2, Will County, Illinois and Docket Nos. STN 50-454 and STN 50-455, Byron Station (Byron), Unit Nos. 1 and 2, Ogle County, Illinois Date of application for amendments: February 23, 2016. Brief description of amendment: The amendments revise technical specifications (TSs) 4.2.1, "Fuel Assemblies," and 5.6.5, "Core Operating Limits Report (COLR)," to allow the use of Optimized ZIRLO TM fuel cladding material in Braidwood, Units 1 and 2, and Byron, Unit Nos. 1 and 2 and to add WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," and Addendum 1-A to Topical Report WCAP-12610-P-A and CENPD-404-P-A, "Optimized ZIRLO" to the list of documents previously reviewed and approved by the NRC. Date of issuance: August 1, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos: 190/196. A publicly-available version is in ADAMS under Accession No. ML16180A251; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: The amendments revised the Technical Specifications and Licenses. Date of initial notice in Federal Register: May 10, 2016 (81 FR 28897). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 1, 2016. No significant hazards consideration comments received: No.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania Date of amendment request: March 24, 2016, as supplemented by letter dated May 11, 2016. Brief description of amendments: The amendments revised the frequency for cycling of the recirculation pump discharge valves as specified in Technical Specification (TS) Surveillance Requirement (SR) 3.5.1.5. Specifically, the amendments changed the frequency for the SR such that it is performed in accordance with the Inservice Testing Program. Date of issuance: August 10, 2016. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendments Nos.: 309 (Unit 2) and 313 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML16165A002; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: June 7, 2016 (81 FR 36619). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 10, 2016. No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant (CNP), Units 1 and 2, Berrien County, Michigan Date of amendment request: January 29, 2016. Brief description of amendments: The amendments revised the CNP, Units 1 and 2, technical specification (TS) requirements to address Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," as described in the Technical Specifications Task Force (TSTF) Traveler, TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation." Date of issuance: August 4, 2016. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. Amendment Nos.: 331 - Unit 1 and 312 - Unit 2. A publicly-available version is in ADAMS under Accession No. ML16195A004; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License Nos. DPR-58 and DPR-74: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register

March 15, 2016 (81 FR 13843). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 4, 2016. No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of amendment request: November 24, 2014, as supplemented by letter dated September 28, 2015. Brief description of amendments: The amendments revised the Technical Specifications (TSs) by adopting 21 previously NRC-approved Technical Specifications Task Force (TSTF) Travelers and one request not associated with TSTF Travelers. SNC stated that these TSTF Travelers are generic changes chosen to increase the consistency between the Joseph M. Farley Nuclear Plant, Units 1 and 2; the Improved Standard Technical Specifications for Westinghouse plants (NUREG-1431); and the TSs of the other plants in the SNC fleet.

Date of issuance: August 3, 2016. Effective date: As of the date of issuance and shall be implemented within 120 days of issuance. Amendment Nos.: 203 (Unit 1) and 199 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML15233A448; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses and TSs. Date of initial notice in Federal Register: February 3, 2015 (80 FR 5804). The supplemental letter dated September 28, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 3, 2016. No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request: March 16, 2016. Brief description of amendments: The amendments revised the Technical Specifications to allow the use of Optimized ZIRLOŽ as an approved fuel rod cladding. Date of issuance: August 4, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 182 (Unit 1) and 163 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16179A386; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: May 24, 2016 (81 FR 32809). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 4, 2016.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of amendment request: March 16, 2016. Brief description of amendments: The amendments revised the Technical Specifications to allow the use of Optimized ZIRLOŽ as an approved fuel rod cladding. Date of issuance: August 4, 2016. Effective date: As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.: 204 (Unit 1) and 200 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML16179A386; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: May 24, 2016 (81 FR 32808). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 4, 2016. No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas Date of amendment request: January 27, 2016, as supplemented by letter dated May 19, 2016. Brief description of amendment: The amendment revised the Technical Specifications to allow the use of Optimized ZIRLO TM as an approved fuel rod cladding. Date of issuance: The amendment is effective upon issuance and shall be implemented within 90 days of the date of issuance.

Effective date: August 3, 2016. Amendment No.: 216. A publicly-available version is in ADAMS under Accession No. ML16179A293; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: April 12, 2016 (81 FR 21603). The supplemental letter dated May 19, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 3, 2016. No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 18 th day of August 2016.

For the Nuclear Regulatory Commission.

/RA/ Anne T. Boland, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

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