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{{Adams
#REDIRECT [[NRC-2016-0273, Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations; Publication Date: January 3, 2017]]
| number = ML16355A261
| issue date = 12/21/2016
| title = Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations; Publication Date: January 3, 2017
| author name = Wilson G A
| author affiliation = NRC/NRR/DORL
| addressee name =
| addressee affiliation =
| docket = 05000255, 05000277, 05000278, 05000302, 05000321, 05000324, 05000325, 05000336, 05000348, 05000364, 05000366, 05000369, 05000370, 05000483
| license number =
| contact person = Burkhardt J C
| case reference number = NRC-2016-0273
| document type = Federal Register Notice
| page count = 39
}}
 
=Text=
{{#Wiki_filter:[7590-01-P]
NUCLEAR REGULATORY COMMISSION [NRC-2016-0273] 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 December 6 to December 19, 2016. The last biweekly notice was published on December 20, 2016.
DATES:  Comments must be filed by February 2, 2017. A request for a hearing must be filed by March 6, 2017.
2 ADDRESSES:  You may submit comments by any of the following methods:
* Federal Rulemaking Web Site:  Go to http://www.regulations.gov and search for Docket ID NRC-2016-0273. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; e-mail:
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:  Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone:
301-415-1384, e-mail:  Janet.Burkhardt@nrc.gov. SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-0273, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of 3
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-2016-0273.
* 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-2016-0273, facility name, unit number(s), plant docket number, application date, and subject in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.
4  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 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 5
amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.
A. Opportunity to Request a Hearing and Petition for Leave to Intervene
 
Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's "Agency Rules of Practice and Procedure" in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued. As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing:  (1) the name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and 6
extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest. In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner must provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, c onsistent with the NRC's regulations, policies, and procedures. Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii).
7 The petition must be filed in accordance with the filing instructions in the "Electronic Submissions (E-Filing)" section of this document. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Co mmission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by March 6, 2017. The petition must be filed in accordance with the filing instructions in the "Electronic Submissions (E-Filing)" section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under
 
10 CFR 2.315(c). If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be 8
permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC Web site
 
at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign 9
submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions should be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an 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 document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system. A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC's Electroni c 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-10 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, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention:  Rulemaking and Adjudications Staff. Participants filing adjudicatory documents in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory 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 personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted 11 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:  December 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16354A424. Description of amendment request:  The amendment would revise Technical Specification (TS) Surveillance Requirement (SR) 4.1.3.1.2 regarding control element assembly (CEA) freedom of movement surveillance, such that CEA 39 may be excluded from the last remaining quarterly performance of the SR in MPS2 Cycle 24. Basis for proposed no significant hazards consideration determination:  As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
 
Response:  No.
 
The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 for the remainder of MPS2 Cycle 24 operation. The function of CEA 39 is to provide negative reactivity addition into the core upon receipt of a signal from the Reactor Protection System (RPS). CEA 39 was demonstrated to 12  be moveable and trippable during the last performance of SR 4.1.3.1.2. Since the functionality of CEA 39 has not been affected, the assumptions and conclusions of the Final Safety Analysis Report (FSAR) Chapter 14, Safety Analysis, are not affected by this license amendment request.
 
The misoperation of a CEA, which includes a CEA drop event, has been evaluated in the MPS2 FSAR and found acceptable. The proposed change would minimize the potential for inadvertent insertion of CEA 39 into the core by eliminating the requirement to place the CEA on the UGC to perform freedom of movement testing. The proposed change does not significantly increase the probability of a failure of a CEA to insert into the core on a reactor trip or the probability of an inadvertent CEA drop into the core at power.
 
No modifications are proposed to the RPS or associated Control Element Drive Mechanism (CEDM) system logic.
 
Based on the above, 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 amendment would exclude CEA 39 from SR 4.1.3.1.2 for the remainder of MPS2 Cycle 24 operation. CEA 39 was demonstrated to be moveable and trippable during the last performance of SR 4.1.3.1.2; therefore, the functionality of CEA 39 has not been affected. The proposed change will not introduce any new design changes or systems that can prevent the CEA from performing its specified safety function to insert on a reactor trip. The current MPS2 FSAR safety analysis considers the drop of a CEA into the core as an initiating event. This change does not alter assumptions made in the FSAR Chapter 14 safety analysis.
 
Based on the above, 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 the margin of safety?
Response:  No.
The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 for the remainder of MPS2 Cycle 24 operation. SR 4.1.3.1.2 is intended to verify freedom of movement of CEAs (i.e., trippable). CEA 39 was 13  demonstrated to be moveable and trippable during the last performance of SR 4.1.3.1.2. The physical and electrical design of the CEAs, and past operating experience, provides high confidence that CEAs remain trippable whether or not exercised during each SR interval. Eliminating further exercise of CEA 39 for the remainder of MPS2 Cycle 24 operation does not directly relate to the potential for CEA binding to occur. The current MPS2 FSAR safety analysis is unaffected by this license amendment request and there is no reduction in the margin of safety.
 
There is no known failure mechanism (e.g., crud deposition) that would preclude the CEA from inserting into core on a valid trip signal or loss of
 
power. Based on the above, the proposed amendment 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:  Lillian M. Cuoco, Seni or Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA  23219. NRC Acting Branch Chief:  Stephen S. Koenick.
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:  June 30, 2016. A publicly-available version is in ADAMS under Accession No. ML16193A656. Description of amendment request:  The amendments would modify Technical Specification 3.6.14, "Divider Barrier Integrity," to revise Conditions A and D to allow one steam generator (SG) enclosure hatch or one pressurizer enclosure hatch to be open for up to 48 hours to facilitate potential inspections and maintenance and to enhance personnel and radiation safety.
14 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.
 
Implementation of this amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. Removal of the SG enclosure hatch or the pressurizer enclosure hatch will not cause an increase in the probability of an accident that has been previously evaluated because the hatches are not accident initiators.
 
The consequences of an accident, which have been previously evaluated, will not be significantly increased by removal of the pressurizer enclosure or SG enclosure hatch. As discussed in the technical justification supporting this amendment request, the new containment compression peak pressure will remain well below the acceptance criteria.
Additionally, the long term containment peak pressure will not be adversely affected due to the delay time in melting of the ice. 
 
The removal of the pressurizer enclosure hatch itself has been previously
 
evaluated in Modes 1 through 4 in accordance with the analytical method described in NUREG-0612 and the NRC's December 22, 1980, letter regarding the control of heavy loads at nuclear power plants [(ADAMS Accession Nos. ML070250180 and ML071080219, respectively)].
Because the SG enclosure hatch weighs less than 300 pounds, it would not be considered a heavy load as defined by NUREG-0612. As such, it is not subject to heavy lift considerations. Regardless, there is no safety-related equipment directly under these hatch covers, so in the unlikely event that one fell, no damage is expected to be caused. The changes proposed in this [license amendment request (LAR)] have no adverse effect on the procedures used for the handling of heavy loads at McGuire.
 
In summary, the proposed changes will not involve any 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.
Operation in accordance with the proposed amendment does not create a new plant configuration and does not adversely affect how the plant is operated, so implementation of this amendment would not create the 15  possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms are created as a result of the NRC approval of this license amendment request. As discussed above, extending the time that the pressurizer enclosure hatch or SG enclosure hatch is allowed to be open or inoperable does not create any new or different accidents from those previously evaluated. Removal of the pressurizer enclosure hatch to perform inspections or maintenance has been previously evaluated and determined to be acceptable. The analysis contained in the technical justification for this license amendment request provides results concluding that the containment compression peak pressure and the long term containment peak pressure are acceptable with either a pressurizer enclosure hatch or [an] SG enclosure hatch open. This proposed amendment does not impact any plant systems that are accident initiators; therefore, no new accident types are being created.
: 3. Does the proposed amendment involve a significant reduction in the margin of safety?
Response:  No.
Implementation of this amendment would not involve a significant reduction in a margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The pressurizer enclosure hatch and the SG enclosure hatch, as well as their performances, have a direct impact on the containment boundary since peak containment pressure due to an accident could be affected. However, the analysis supporting this amendment request concludes that the containment compression peak pressure and the long term containment peak pressure continue to be acceptable with the increased time a single hatch is open. 
 
Therefore, the performance of the fission product barriers will not be significantly impacted by implementation of this amendment, and no safety margins will be significantly impacted. 
 
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:  Kate Nolan, Deputy General Counsel, Duke Energy Corporation, 526 South Church Street - DEC45A, Charlotte, NC  28202.
16 NRC Branch Chief:  Michael T. Markley. 
 
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North Carolina Date of amendment request:  September 6, 2016, as supplemented by letter dated November 9, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16257A410 and ML16330A504, respectively.
Description of amendment request:  The amendment would revise the Technical Specifications to support an expansion of the core power-flow operating range (i.e., Maximum Extended Load Line Limit Analysis Plus (MELLLA+)).
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 operation in the MELLLA+ operating domain does not significantly increase the probability or consequences of an accident previously evaluated. The probability (i.e., frequency of occurrence) of Design Basis Accidents (DBAs) occurring is not affected by the MELLLA+
operating domain because BSEP continues to comply with the regulatory and design basis criteria established for plant equipment. Furthermore, a probabilistic risk assessment demonstrates that the calculated core damage frequencies do not significantly change due to the MELLLA+.
There is no change in consequences of postulated accidents when operating in the MELLLA+ operating domain compared to the operating domain previously evaluated. The results of accident evaluations remain within the NRC approved acceptance limits. 
 
The spectrum of postulated transients has been investigated and is shown to meet the plant's currently licensed regulatory criteria.
17  Continued compliance with the Safety Limit Minimum Critical Power Ratio (SLMCPR) will be confirmed on a cycle-specific basis consistent with the criteria accepted by the NRC. 
 
Challenges to the reactor coolant pressure boundary were evaluated for the MELLLA+ operating domain conditions (i.e., pressure, temperature, flow, and radiation) and were found to meet their respective acceptance criteria for allowable stresses and overpressure margin. 
 
Challenges to the containment were evaluated and the containment and its associated cooling systems cont inue to meet the current licensing basis. The calculated post-Loss of Coolant Accident (LOCA) suppression pool temperature remains acceptable. 
 
The proposed changes to the sodium pentaborate (SPB) enrichment and volume requirements maintain the capability of the Standby Liquid Control (SLC) system to perform this reactivity control function and ensure continued compliance with the requirements of 10 CFR 50.62. The SLC system is not considered to be an initiator of any event. The use of the proposed SPB solution with a higher boron-10 (B-10) isotope enrichment does not alter the design, function, or operation of the SLC system or increase the likelihood of malfunction that could increase the consequences of an accident.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:  No.
 
The proposed operation in the MELLLA+ operating domain does not create the possibility of a new or different kind of accident from any previously evaluated.
 
Equipment that could be affected by the MELLLA+ operating domain has
 
been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified. The full spectrum of accident considerations has been evaluated and no new or different kind of accident was identified. The MELLLA+ operating domain uses developed technology, and applies it within the capabilities of existing plant safety-related equipment in accordance with the regulatory criteria, including NRC-approved codes, standards and methods. The use of the proposed SPB solution with a higher B-10 isotope enrichment does not alter the design, function, or operation of the SLC system or create the possibility of a new or different kind of accident.
The proposed changes have been assessed and determined not to 18  introduce a different accident than that previously evaluated. No new accident or event precursor has been identified.
 
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 operation in the MELLLA+ domain does not involve a significant reduction in the margin of safety.
 
The MELLLA+ operating domain can only affect design and operational margins. Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for the MELLLA+ operating domain conditions. Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on affected structures, systems, and components, including the reactor coolant pressure boundary, will remain within their design allowables for design basis event categories.
No NRC acceptance criterion is exceeded. The BSEP configuration and responses to transients and postulated accidents do not result in exceeding the presently approved NRC acceptance limits, thereby preserving safety margins. 
 
The proposed changes to the SPB enrichment and volume requirements ensure SLC system shutdown margins and post-accident pH control margins are maintained while maintaining compliance with the requirements of 10 CFR 50.62. 
 
Therefore, the proposed amendments do not result in 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:  Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon St., M/C DEC45A, Charlotte, NC  28202.
NRC Acting Branch Chief:  Jeanne D. Johnston.
19 Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3 (PBAPS), York and Lancaster Counties, Pennsylvania Date of amendment request:  November 4, 2016, as supplemented by letter dated December 7, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16309A298 and ML16342C455, respectively.
Description of amendment request:  The amendments w ould revise the Allowable Value (AV) for the Turbine Condenser - Low Vacuum scram function specified in Technical Specification Table 3.3.1.1-1, "Reactor Protection System Instrumentation."  The licensee stated that the purpose of the proposed change is to minimize the potential for inadvertent scrams due to low condenser vacuum. 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 edits shown in square brackets: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:  No. 
 
The proposed change implements a revised AV [allowable value] for the Turbine Condenser - Low Vacuum scram instrument function at PBAPS.
The proposed change to the PBAPS Turbine Condenser- Low Vacuum scram AV does not require modifying any system interface or affect the probability of any event initiators at the facility. Overall RPS [Reactor Protection System] performance will remain within the bounds of the previously performed accident analyses, since [the Turbine-Condenser - Low Vacuum scram is not specifically credited in any accident analysis.]
 
There will be no degradation in the performance of, or an increase in the
 
number of challenges imposed on safety-related equipment that are 20  assumed to function during an accident situation. The proposed change will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the UFSAR [Updated Final Safety Analysis Report]. The proposed change is consistent with safety analysis assumptions and resultant consequences.
 
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:  No. 
 
The proposed change to the PBAPS Turbine Condenser - Low Vacuum scram AV does not affect the design, functional performance, or operation of the facility. Similarly, the proposed change does not affect the design or operation of any SSCs [structures, systems, or components] involved in the mitigation of any accidents, nor does it affect the design or operation of any component in the facility such that new equipment failure
 
modes are created.
 
No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. There will be no adverse effect or challenges imposed on any safety-related system as a result of this 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.
The proposed change does not affect the acceptance criteria for any analyzed event, nor is there a change to any Safety Analysis Limit. There will be no effect on the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. 
 
The purpose of the condenser low vacuum turbine trip is to protect the main condenser against overpressure on loss of condenser vacuum. A condenser low vacuum condition provides a signal to trip the main turbine by providing automatic closure to the turbine stop valves. To anticipate the transient and scram which results from the closure of the turbine stop valves, a condenser low vacuum condition initiates a reactor scram. The 21  condenser low vacuum scram trip setting is selected to initiate a reactor scram prior to initiation of closure of the [t]urbine [s]top [v]alves.
 
The proposed LAR [license amendment request] does not change the sequential relationship of the condenser low vacuum scram and turbine trip. The Automatic Scram signal (Actual Trip Setpoint greater than or equal to 21.95 inches [mercury (Hg)] vacuum) will still occur prior to the Turbine Trip signal (Actual Trip Setpoint 20.0 inches Hg vacuum). This aligns with UFSAR Section 7.2 in that the condenser low vacuum scram is an anticipatory trip prior to the scram that would result from the closure of the main turbine stop valves. 
 
The condenser low vacuum scram is not specifically credited in any accident analysis. The integrity of the condenser is not compromised by the proposed change because the reactor will be shut down using both diverse and redundant tripping to ensure fission products are not
 
released.
Therefore, the proposed change does not involve a significant reduction
 
in a margin of safety. 
 
The NRC staff has reviewed the licensee's analysis. Based on this review, and the NRC edits shown in square brackets, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee:  Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL  60555. NRC Acting Branch Chief:  Stephen S. Koenick. 
 
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 22, 2016. A publicly-available version is in ADAMS under Accession No. ML16336A024.
22 Description of amendment request:  SNC requested to revise the licensing basis that support a selected scope application of an Alternative Source Term (AST) methodology and incorporate Technical Specification Task Force (TSTF) Traveler, TSTF-448-A, Revision 3, "Control Room
 
Habitability," and TSTF-312-A, "Administrativ e Control of Containment Penetrations." Basis for proposed no significant hazards consideration determination:  As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The licensee's analysis is presented below, with NRC staff edits in square brackets: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:  No.
 
There are no physical changes to the plant being introduced by the proposed changes to the accident source term. Implementation of Alternative Source Term (AST) and the new atmospheric dispersion factors have no impact on the probability for initiation of any Design Basis Accidents (DBAs). Once the occurrence of an accident has been postulated, the new accident source term and atmospheric dispersion factors are an input to analyses that evaluate the radiological consequences. The proposed changes do not involve a revision to the design or manner in which the facility is operated that could increase the probability of an accident previously evaluated in Chapter 15 of the Final Safety Analysis Report (FSAR).
 
Based on the AST analyses, there are no proposed changes to performance requirements and no proposed revision to the parameters or conditions that could contribute to the consequences of an accident previously discussed in Chapter 15 of the FSAR. Plant-specific radiological analyses have been performed using the AST methodology and new atmospheric dispersion factors (X/Qs) have been established. Based on the results of these analyses, it has been demonstrated that the Control Room and off-site dose consequences of the limiting events considered in the analyses meet the regulatory guidance provided for use with the AST, and the doses are within the limits established by 10 CFR
 
50.67. Regarding TSTF-312-A, the proposed change would allow containment penetrations to be unisolated under administrative controls during core 23  alterations or movement of irradiated fuel assemblies within containment. The status of containment penetration flow paths (i.e., open or closed) is not an initiator for any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The proposed change does not affect the design of the primary containment, or alter plant operating practices such that the probability of an accident previously evaluated would be significantly increased. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated, and is bounded by the fuel handling accident (FHA) analysis.
 
Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or the 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.
No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents. Implementation of AST and the associated proposed Technical Specification changes and new X/Qs have no impact to the initiation of any DBAs. These changes do not affect the design function or modes of operation of structures, systems and components in the facility prior to a postulated accident. Since structures, systems and components are operated no differently after the AST implementation, no new failure modes are created by this proposed change. The AST change itself does not have the capability to initiate accidents.
 
Regarding TSTF-312-A, allowing penetration flow paths to be open is not an initiator for any accident. The proposed change to allow open penetration flow paths will not affect plant safety functions or plant operating practices such that a new or different accident could be created. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. 
 
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
 
Consequently, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
 
24  3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response:  No.
The AST analyses have been performed using approved methodologies to ensure that analyzed events are bounding and safety margin has not been reduced. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67. Thus, by meeting the applicable regulatory limits for AST, there is no significant reduction in a margin of safety.
 
Regarding TSTF-312-A, TS 3.9.3 provides measures to ensure that the dose consequences of a postulated FHA inside containment are minimized. The proposed change to LCO 3.9.3 will allow penetration flow path(s) to be open during refueling operations under administrative control. These administrative controls will provide assurance that prompt closure of open penetrations flow paths can and will be achieved in the event of an FHA inside containment, and will minimize dose consequences. The proposed change does not affect the safety analysis acceptance criteria for ay [an] analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the
 
manner in which safety limits, limiting safety system settings or limiting conditions for operation are deterred, not [nor] is here [there] any adverse effect on those plant systems necessary to assure the accomplishment of protective functions. The proposed change will not result in plant operation in a configuration outside the design basis.
 
Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the proposed amendment does not involve a significant reduction in 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, 40 Inverness Center Parkway, Birmingham, AL  35201.
NRC Branch Chief:  Michael T. Markley. 
 
25 Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of amendment request:  November 15, 2016. A publicly-available version is in ADAMS under Accession No. ML16320A540. Description of amendment request:  The proposed changes would revise Technical Specification 5.5.17, "Containment Leakage Rate Testing Program."  The revision would increase the existing Type A integrated leakage rate test program test interval from 10 years to 15 years; adopt an extension of the containment isolation valve leakage testing (Type C) frequency from 60 months to 75 months; adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, "Containment System Leakage Testing Requirements"; and adopt a grace interval of 9 months for Type A, Type B, and Type C leakage tests, in accordance with Nuclear Energy Institute (NEI) 94-01, Revision 3-A. 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 activity involves the revision of Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Technical Specification (TS) 5.5.17, "Primary Containment Leakage Rate Testing Program," to allow the extension of the Type A integrated leakage rate test (ILRT) containment test interval to 15 years, and the extension of the Type C local leakage rate test (LLRT) interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions.
 
26  The proposed extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the
 
testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.
 
The change in Type A test frequency to once-per-fifteen years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, based on the internal events probabilistic risk analysis (PRA) is 1.08E-02 person-rem/year for Unit 1 and 9.89 E-03 person-rem/year for Unit 2. Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2-A states that a very small population is defined as an increase of  1.0 person-rem per year or  1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This is consistent with the Nuclear Regulatory Commission (NRC) Final Safety Evaluation for Nuclear Energy Institute (NEI) 94-01 and EPRI Report No. 1009325.
Moreover, the risk impact when compar ed to other severe accident risks is negligible.
Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.
 
In addition, as documented in NUREG-1493, "Performance-Based Containment Leak-Test Program," dated January 1995, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The FNP Type A test history supports this conclusion.
 
The integrity of the containment is subject to two types of failure mechanisms that can be categorized as:  (1) activity based, and (2) time based. Activity-based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. The LLRT requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME)
Section XI, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated.
27  The proposed amendment also deletes exceptions previously granted under TS Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to allow one-time extensions of the ILRT test frequency for FNP. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.
 
Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:  No.
The proposed amendment to the TS 5.5.17, "Containment Leakage Rate Testing Program," involves the extension of the FNP Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.
The proposed amendment also deletes exceptions previously granted under TS Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to allow one-time extensions of the ILRT test frequency for FNP. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that does not result in any change in how the unit is
 
operated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?  Response:  No.
The proposed amendment to TS 5.5.17 involves the extension of the FNP
 
Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are 28  determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.
 
The proposed change involves only the extension of the interval between Type A containment leak rate tests and Type C tests for FNP. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusions that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI and Technical Specifications serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.
 
The proposed amendment also deletes an exception previously granted under TS Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to allow one-time extensions of the ILRT test frequency for FNP. This exception was for an activity that would have already taken place by the time this amendment is approved; therefore, the deletion is solely an administrative action and does not change how the unit is operated and maintained.
Therefore, there is no reduction in any margin of 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:  Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center Parkway, Birmingham, AL  35242.
NRC Branch Chief:  Michael T. Markley.
29 Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri Date of amendment request:  October 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16286A553. Description of amendment request:  The amendment would revise the Technical Specification (TS) requirements to reference and allow use of the NRC-approved core reload methodologies described in Westinghouse topical reports (TRs) WCAP-16045-P-A, "Qualification of the Two-
 
Dimensional Transport Code PARAGON," WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology," and WCAP-10965-P-A, Addendum 2-A, "Qualification of the New Pin Power Recovery Methodology," for the Callaway Plant.
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 license amendment would revise TS 5.6.5.b to add additional TR references for NRC-approved methodologies used in core reload designs and the determination of core operating limits, thereby specifically approving the use of these methodologies for the Callaway Plant. The additional analytical methodologies are improvements over the current methodologies in use at Callaway Plant. The NRC staff reviewed and approved these methodologies and concluded that these analytical methods are acceptable as a replacement for the current
 
analytical method. 
 
This proposed license amendment does not involve any physical changes to the Callaway Plant. Additionally, the core operating limits determined using the proposed analytical methods will continue to assure that the reactor operates safely. On that basis, the proposed changes do not involve an increase in the probability of an accident. 
 
30  The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the
 
assumed acceptance limits and therefore, does not increase the likelihood of any failure mechanisms or precursors to transients or accidents postulated and analyzed in the Callaway Plant FSAR [Final Safety Analysis Report]. Operation of the reactor with core operating limits determined by use of the proposed analytical methods does not increase the reactor power level, does not increase the core fission product inventory, and does not change any radiological release assumptions. The proposed changes will not alter any accident analysis assumptions discussed in the FSAR, nor do they involve any changes to the requirement for Callaway Plant to operate within the power distribution limits and shutdown margins required by the TS and within the assumptions of the safety analyses described in the FSAR. Therefore the proposed methodology and TS changes do not involve a significant increase in the consequences of an accident.
Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously
 
evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
 
Response:  No.
 
The proposed change provides revised analytical methods for determining core operating limits, and does not change any system functions or requirements. Acceptance criteria required to be met for analyzed core performance under normal, transient and accident conditions are not being changed, as the core operating limits will continue to be established in accordance with NRC-approved methods.
The change does not involve physical alteration of the plant, as no new or different type of equipment will be installed. The change does not alter assumptions made in the safety analyses, but ensures that the core will operate within safe limits. Consequently, this change does not create new failure modes or mechanisms, and no new accident precursors are
 
generated.
Therefore, it is concluded that this change does not create the possibility 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.
 
31  The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes do not physically alter safety-related systems, nor do they affect the way in which safety related systems perform their functions. The setpoints at which protective actions are initiated are not altered by the proposed changes. The availability of equipment required to be available to actuate upon demand for mitigating an analyzed event is unchanged by the proposed amendment. The proposed analytical methodologies are an improvement that allows more accurate modeling of core performance. The NRC has reviewed and approved the additional methodologies for use in lieu of the current methodology; thus, the margin of safety is not reduced due to this change.
Therefore, it is concluded that 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:  John O'Neill, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, N.W., Washington, DC  20037.
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 32 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.
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:  September 26, 2013, as supplemented by letters dated January 8, 2014; October 13, 2014; November 12, 2014; December 12, 2014; January 26, 2015; February 27, 2015; March 13, 2015; July 15, 2015; August 20, 2015; September 9, 2015; October 1, 2015; January 14, 2016; April 26, 2016; September 29, 2016; and November 21, 2016.
33 Brief description of amendments:  The amendments revised the condition for the fire protection program (FPP) in Facility Operating Licenses such that the FPP is now based on the requirements of 10 CFR 50.48(c), "National Fire Protection Association Standard NFPA 805."
Date of issuance:  December 6, 2016. Effective date:  As of the date of issuance and shall be implemented as stated in the revised License Condition 2.C(4). Amendment Nos.:  291 and 270. A publicly-available version is in ADAMS under Accession No. ML16077A135; 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. Date of initial notice in Federal Register:  February 19, 2014 (79 FR 9492). The supplemental letters dated January 8, 2014; October 13, 2014; November 12, 2014; December 12, 2014; January 26, 2015; February 27, 2015; March 13, 2015; July 15, 2015; August 20, 2015; September 9, 2015; October 1, 2015; January 14, 2016; April 26, 2016; September 29, 2016; and November 21, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposal 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 December 6, 2016. No significant hazards consideration comments received:  No. 
 
34 Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida Date of application for amendment:  July 28, 2015. Brief description of amendment:  The amendment incorporates into the license the transfer of ownership, held by Seminole Electric Cooperative, Inc. (SEC), in CR-3 to DEF. The transfer of ownership will take place pursuant to the Settlement, Release and Acquisition Agreement, dated April 30, 2015, wherein DEF will purchase the 1.6994 percent ownership share in CR-3 held by SEC, leaving DEF as the sole remaining licensee for CR-3.
Date of issuance:  November 30, 2016. Effective date:  As of the date of issuance and shall be implemented within 60 days. Amendment No.:  251. A publicly-available version is in ADAMS under Accession No. ML16293A200; documents related to this amendment are listed in the Safety Evaluation enclosed with the letter dated August 10, 2016 (ADAMS Accession No. ML16173A022). Facility Operating License No. DPR-72:  This amendment revised the Facility Operating License.
Date of initial notice in Federal Register:  September 29, 2015 (80 FR 58513), and January 4, 2016 (81 FR 98). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 10, 2016.
No significant hazards consideration comments received:  No. 
 
Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida 35 Date of application for amendment:  August 27, 2015, as supplemented by letters dated March 2, 2016, and July 14, 2016. Brief description of amendment:  The amendment approved the CR-3 Permanently Defueled Emergency Plan, and Permanently Defueled Emergency Action Level Bases Manual, for the Independent Spent Fuel Storage Installation. Date of issuance: December 5, 2016. Effective date:  As of the date of issuance and shall be implemented within 60 days. Amendment No.:  252. A publicly-available version is in ADAMS under Accession No. ML16244A099; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Facility Operating License No. DPR-72:  This amendment revised the License. Date of initial notice in Federal Register
:  November 10, 2015 (80 FR 69711). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 5, 2016. No significant hazards consideration comments received:  No. 
 
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant, Van Buren County, Michigan Date of amendment request:  March 3, 2016, as supplemented by letter dated June 7, 2016. Brief description of amendment:  The amendment approves the implementation of an alternate repair criteria (ARC) called C-star, for the portion of the steam generator (SG) tubes within the cold-leg tubesheet. In addition, the amendment clarifies the intent and improves the wording of the technical specifications regarding the previously incorporated ARC for the hot-leg side of the 36 SG's tubesheet. This was previously approved by letter dated May 31, 2007, and Amendment No. 225. Date of issuance:  December 19, 2016. Effective date:  As of the date of issuance and shall be implemented within 60 days. Amendment No.:  261. A publicly-available version is in ADAMS under Accession No. ML16300A030; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment. Renewed Facility Operating License No. DPR-20:  Amendment revised the Renewed Facility Operating License and Technical Specifications. Date of initial notice in Federal Register:  August 2, 2016 (81 FR 50747). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 19, 2016. No significant hazards consideration comments received:  No.
Southern Nuclear Operating Company, Inc.; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia Date of amendment request:  December 15, 2015, as supplemented by letter dated April 11, 2016. Brief description of amendments:  The amendments revised the Technical Specifications (TSs) to incorporate risk-informed requirements regarding selected Required Action end states. Additionally, it modified TS Required Actions with a Note prohibiting the use of Limiting Condition for Operation Applicability 3.0.4.a when entering the preferred end state (Mode 3). Date of issuance:  December 19, 2016.
37 Effective date:  As of the date of issuance and shall be implemented within 90 days of issuance. Amendment Nos.:  281 (Unit No. 1); 225 (Unit No.
2). A publicly-available version is in ADAMS under Accession No. ML16257A724; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments. Renewed Facility Operating License Nos. DPR-57 and NPF-5:  Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register:  February 16, 2016 (81 FR 7841). The supplemental letter dated April 11, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 19, 2016. No significant hazards consideration comments received:  No. 
 
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee Date of amendment request:  February 23, 2016, as supplemented by letter dated July 22, 2016. Brief description of amendment:  The amendment approved revisions to the WBN Dual Unit Fire Protection Report and revised the associated License Condition regarding the WBN fire protection program. Date of issuance:  December 12, 2016. Effective date:  As of the date of issuance and shall be implemented within 30 days of issuance.
38 Amendment No.:  108. A publicly-available version is in ADAMS under Accession No. ML16307A013; 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:  May 10, 2016 (81 FR 28901). The supplemental letter dated July 22, 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.
39  The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 12, 2016. No significant hazards consideration comments received:  No. 
 
Dated at Rockville, Maryland, this 21 st day of December 2016.
For the Nuclear Regulatory Commission. 
 
/RA/
George A. Wilson, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.}}

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