81 Fed. Reg. 87964 (Dec. 6, 2016)

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

Download PDF

Federal RegisterDec 6, 2016

81 Fed. Reg. 87964 (Dec. 6, 2016)

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued, from November 8 to November 21, 2016. The last biweekly notice was published on November 22, 2016.

DATES:

Comments must be filed by January 5, 2017. A request for a hearing must be filed by February 6, 2017.

ADDRESSES:

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

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0245. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1384, email: Janet.Burkhardt@nrc.gov.

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2016-0245, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0245.
NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/ adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2016-0245, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

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

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

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

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

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

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

Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and a petition to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel will issue a notice of a hearing or an appropriate order.

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

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

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions consistent with the NRC's regulations, policies, and procedures.

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

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

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

The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by February 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. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c).

If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene (hereinafter “petition”), and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

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

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

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

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC Electronic Filing Help Desk through the “Contact Us” link located on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html,, by email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 7 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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

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

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

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

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

Date of amendment request: August 29, 2016. A publicly-available version is in ADAMS under Accession No. ML16252A220.

Description of amendment request: The amendments would revise the Technical Specifications (TSs) to eliminate Section 5.5, “lnservice Testing Program.” A new defined term, “lnservice Testing Program,” is added to the TS Definitions section. This request is consistent with Technical Specification Task Force Traveler TSTF-545, Revision 3, “TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing” (ADAMS Accession No. ML15314A305).

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

Response: No.

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

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

Response: No.

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

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

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

Response: No.

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

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

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

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

Acting NRC Branch Chief: Jeanne A. Dion.

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

Date of amendment request: September 28, 2016. A publicly-available version is in ADAMS under Accession No. ML16287A415.

Description of amendment request: The amendments would revise the Technical Specifications (TSs) to be consistent with Technical Specification Task Force Traveler TSTF-423, Revision 1, to allow, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Changes proposed in TSTF-423 will be made to the Units' TSs for selected Required Action end states.

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

Response: No.

The proposed change allows a change to certain required end states when the TS Completion Times for remaining in power operation will be exceeded. Most of the requested TS changes are to permit an end state of hot shutdown (Mode 3) rather than an end state of cold shutdown (Mode 4) contained in the current TS. The request was limited to: (1) those end states where entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical. Risk insights from both the qualitative and quantitative risk assessments were used in specific TS assessments. Such assessments are documented in Section 6 of topical report NEDC-32988-A, Revision 2, “Technical Justification to Support Risk-Informed Modification to Selected Required Action End States for BWR [Boiling-Water Reactor] Plants.” They provide an integrated discussion of deterministic and probabilistic issues, focusing on specific TSs, which are used to support the proposed TS end state and associated restrictions. The NRC staff finds that the risk insights support the conclusions of the specific TS assessments. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting TSTF-423 are no different than the consequences of an accident prior to adopting TSTF-423. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). If risk is assessed and managed, allowing a change to certain required end states when the TS Completion Times for remaining in power operation are exceeded (i.e., entry into hot shutdown rather than cold shutdown to repair equipment) will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change and the commitment by the licensee to adhere to the guidance in TSTF-IG-05-02, “Implementation Guidance for TSTF-423, Revision 1, `Technical Specifications End States, NEDC-32988-A,'” will further minimize possible concerns.

Thus, based on the above, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

Response: No.

The proposed change allows, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed. The BWROG's [BWR Owner Group's] risk assessment approach is comprehensive and follows NRC staff guidance as documented in Regulatory Guides (RG) 1.174 and 1.177. In addition, the analyses show that the criteria of the three-tiered approach for allowing TS changes are met. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A risk assessment was performed to justify the proposed TS changes. The net change to the margin of safety is insignificant.

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

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

Acting NRC Branch Chief: Jeanne A. Dion.

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

Date of amendment request: August 30, 2016. A publicly-available version is in ADAMS under Accession No. ML16245A273.

Description of amendment request: The amendment would reclassify reactor water cleanup (RWCU) piping, valves, pumps and mechanical modules located outside of primary and secondary containment in the radwaste building from Quality Group C to Quality Group D.

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

Response: No.

The proposed amendment does not result in a significant increase in the probability of an accident because Quality Group D standards are considered appropriate for water containing components which are not part of the reactor coolant pressure boundary but may contain radioactive materials. The probability of a line break is not increased since the materials, design, and fabrication of Quality Group C components is comparable to Quality Group D components. Differences between the two quality groups are limited primarily to quality assurance requirements. The use of Quality Group D components for portions of RWCU located in the radwaste building provides an adequate level of quality, commensurate with the importance of the functions to be performed by that portion of the system, and ensures that the facility can be operated without undue risk to the health and safety of the public.

All safety related equipment required to mitigate accidents is either significantly remote from, or separated by protective barriers from the reclassified portions of the system. The consequences of breaks considered in the portion of the RWCU system affected by this activity are calculated to not exceed regulatory limits for dose to control room personnel or the public. Calculated results are not significantly different than those reported for the existing instrument line break analysis in [the Final Safety Analysis Report (FSAR)] Chapter 15.

[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.

A postulated failure in the RWCU system piping would result in a high-energy line break (HELB) accident. High energy line breaks are already postulated and analyzed at various locations for portions of the RWCU system located in the reactor building. The existing instrument line break analysis was determined to bound a postulated worst case RWCU HELB. Since the offsite and onsite consequences of a postulated break in the reclassified portion of the RWCU is bounded by the existing instrument line break analyses, a new or different accident has not been created.

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

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

Response: No.

The proposed amendment does not involve a significant reduction in a margin of safety. As noted in the technical and regulatory evaluation above, the reclassified portions of the system perform no active safety functions and will not result in radiological safety impact beyond that already assumed within the existing plant safety analyses.

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

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

NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Date of amendment request: October 7, 2016. A publicly-available version is in ADAMS under Accession No. ML16281A174.

Description of amendment request: The amendments would revise the Byron Station licensing basis for protection from tornado-generated missiles. Specifically, the Updated Final Safety Analysis Report (UFSAR) would be revised to identify the TORMIS Computer Code as the methodology used for assessing tornado-generated missile protection of unprotected plant structures, systems, and components (SSCs) and to describe the results of the Byron Station site-specific tornado hazard analysis.

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

Response: No.

The NRC TORMIS Safety Evaluation Report states the following:

“The current Licensing criteria governing tornado missile protection are contained in [NUREG-0800] Standard Review Plan (SRP) Section 3.5.1.4, [Missiles Generated by Natural Phenomena] and 3.5.2 [Structures, Systems and Components to be Protected from Externally Generated Missiles]. These criteria generally specify that safety-related systems be provided positive tornado missile protection (barriers) from the maximum credible tornado threat. However, SRP Section 3.5.1.4 includes acceptance criteria permitting relaxation of the above deterministic guidance, if it can be demonstrated that the probability of damage to unprotected essential safety-related features is sufficiently small.”

As permitted by these SRP sections, the combined probability will be maintained below an allowable level, i.e., an acceptance criterion threshold, which reflects an extremely low probability of occurrence. SRP Section 2.2.3, “Evaluation of Potential Accidents,” established this threshold as approximately 1.0E-06 per year if, when combined with reasonable qualitative arguments, the realistic probability can be shown to be lower. The Byron Station analysis approach assumes that if the sum of the individual probabilities calculated for tornado missiles striking and damaging portions of safety-significant SSCs is greater than or equal to 1.0E-06 per year per unit, then installation of tornado missile protection barriers would be required for certain components to lower the total cumulative damage probability below the acceptance criterion of 1.0E-06 per year per unit. Conversely, if the total cumulative damage probability remains below the acceptance criterion of 1.0E-06 per year per unit, no additional tornado missile protection barriers would be required for any of the unprotected safety-significant components.

With respect to the probability of occurrence or the consequences of an accident previously evaluated in the UFSAR, the possibility of a tornado impacting the Byron Station site and causing damage to plant SSCs is a licensing basis event currently addressed in the UFSAR. The change being proposed (i.e., the use of the TORMIS methodology for assessing tornado-generated missile protection of unprotected plant SSCs), does not affect the probability of a tornado strike on the site; however, from a licensing basis perspective, the proposed change does affect the probability that missiles generated by a tornado will strike and damage certain safety-significant plant SSCs. There are a defined number of safety-significant components that could theoretically be struck and damaged by tornado-generated missiles. The probability of tornado-generated missile hits on these “important” systems and components is calculated using the TORMIS probabilistic methodology. The combined probability of damage for unprotected safety-significant equipment will be maintained below the acceptance criterion of 1.0E-06 per year per unit to ensure adequate equipment remains available to safely shutdown the reactors, and maintain overall plant safety, should a tornado strike occur. Consequently, the proposed change does not constitute a significant increase in the probability of occurrence or the consequences of an accident based on the extremely low probability of damage caused by tornado-generated missiles and the commensurate extremely low probability of a radiological release.

Finally, the use of the TORMIS methodology will have no impact on accident initiators or precursors; does not alter the accident analysis assumptions or the manner in which the plant is operated or maintained; and does not affect the probability of operator error.

Based on the above discussion, 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 impact of a tornado strike on the Byron Station site is a licensing basis event that is explicitly addressed in the UFSAR. The proposed change simply involves recognition of the acceptability of using an analysis tool (i.e., the TORMIS methodology) to perform probabilistic tornado missile damage calculations in accordance with approved regulatory guidance. The proposed change does not result in the creation of any new accident precursors; does not result in changes to any existing accident scenarios; and does not introduce any operational changes or mechanisms that would create the possibility of a new or different kind of accident.

Therefore, the proposed change will not create the possibility of a new or different kind of accident than those previously evaluated.

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

Response: No.

The existing Byron Station licensing basis regarding tornado missile protection of safety-significant SSCs assumes that missile protection barriers are provided for safety-significant SSCs; or the unprotected component is assumed to be unavailable post-tornado. The results of the Byron Station TORMIS analysis have demonstrated that there is an extremely low probability, below an established regulatory acceptance limit, that these “important” SSCs could be struck and subsequently damaged by tornado-generated missiles. The change in licensing basis from protecting safety-significant SSCs from tornado missiles, to demonstrating that there is an extremely low probability that safety-significant SSCs will be struck and damaged by tornado-generated missiles, does not constitute a significant decrease in the margin of safety.

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

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

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

Acting NRC Branch Chief: G. Edward Miller.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station (CPS), Unit 1, DeWitt County, Illinois

Date of amendment request: August 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16229A278.

Description of amendment request: The proposed change would eliminate the on-shift positions not needed for storage of the spent fuel in the spent fuel pool during the initial decommissioning period and the emergency response organization (ERO) positions not needed to respond to credible events. Additionally the licensee is proposing to revise the emergency action levels (EALs) to reflect those conditions applicable when the unit is in a permanently defueled condition.

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

Response: No.

The proposed changes to the CPS Emergency Plan do not impact the function of plant Structures, Systems, or Components (SSCs). The proposed changes do not involve the modification of any plant equipment or affect plant operation. The proposed changes do not affect accident initiators or precursors, nor do the proposed changes alter design assumptions. The proposed changes do not prevent the ability of the on-shift staff and ERO to perform their intended functions to mitigate the consequences of any accident or event that will be credible in the permanently defueled condition. The proposed changes only remove positions and remove certain EALs that will no longer be needed or credited in the Emergency Plan in the permanently defueled condition.

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

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

Response: No.

The proposed changes reduce the number of on-shift and ERO positions commensurate with the hazards associated with a permanently shutdown and defueled facility. The proposed changes also remove EALs which are no longer applicable to CPS in a permanently defueled condition. The proposed changes do not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed changes do not result in a change to the way that the equipment or facility is operated so that no new accident initiators are created.

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

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

Response: No.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. The proposed changes are associated with the Emergency Plan and staffing and EAL schemes and do not impact operation of the plant or its response to transients or accidents.

The proposed changes do not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by the proposed changes and margins of safety are maintained. The revised Emergency Plan will continue to provide the necessary response staff with the proposed changes.

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

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

Acting NRC Branch Chief: G. Edward Miller.

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

Date of amendment request: October 20, 2016. A publicly-available version is in ADAMS under Accession No. ML16294A203.

Description of amendment request: The proposed amendment request supports the deletion, modification, and addition to the organization, staffing, and training requirements contained in Sections 1.0 and 5.0 of the Technical Specifications (TSs) after the license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor pressure vessel. This proposed amendment also supports implementation of the Certified Fuel Handler training program.

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

Response: No.

The proposed changes would not take effect until QCNPS has permanently ceased operation and entered a permanently defueled condition. The proposed changes would revise the QCNPS TS by deleting or modifying certain portions of the TS administrative controls described in Section 5.0 of the TS that are no longer applicable to a permanently shutdown and defueled facility.

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

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

In a permanently defueled condition, the only credible accidents are the Design Basis Fuel Handling Accidents Inside Containment (the specific concern is dropping a fuel bundle over the Spent Fuel Pool; not the Reactor Vessel) and Spent Fuel Storage Buildings and Postulated Liquid Releases Due to Liquid Tank Failures. Other accidents such as Loss of Coolant Accident, Loss of Feedwater, and Reactivity and Power Distribution Anomalies will no longer be applicable to a permanently defueled reactor plant.

The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a permanently defueled condition will be the only operation allowed, and therefore, bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

The proposed changes in the administrative controls do not affect the ability to successfully respond to previously evaluated accidents and do not affect radiological assumptions used in the evaluations. The proposed changes narrow the focus of nuclear safety concerns to those associated with safely maintaining spent nuclear fuel. These changes remove the implication that QCNPS can return to operation once the final certification required by 10 CFR 50.82(a)(1)(ii) is submitted to the NRC. Any event involving safe storage of spent irradiated fuel or the methods used for handling and storage of such fuel in the SFP would evolve slowly enough that no immediate response would be required to protect the health and safety of the public or station personnel. Adequate communications capability is provided to allow facility personnel to safely manage storage and handling of irradiated fuel. As a result, no changes to radiological release parameters are involved. There is no effect on the type or amount of radiation released, and there is no effect on predicted offsite doses in the event of an accident.

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

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

Response: No.

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

The proposed changes do not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers in support of maintaining the plant in a permanently shutdown and defueled condition (e.g., fuel cladding and SFP cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

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

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

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

Response: No.

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

The only remaining credible accidents are the Design Basis Fuel Handling Accidents Inside Containment and Spent Fuel Storage Buildings (the specific concern is dropping a fuel bundle over the Spent Fuel Pool; not the Reactor Vessel) and the Postulated Liquid Releases Due to Liquid Tank Failures. The proposed changes do not adversely affect the inputs or assumptions of any of the design basis analyses that impact the Design Basis Fuel Handling Accidents.

The proposed changes are limited to those portions of the TS administrative controls that are not related to the safe storage and maintenance of spent irradiated fuel.

These proposed changes do not directly involve any physical equipment limits or parameters. The requirements that are proposed to be revised and/or deleted from the QCNPS TS are not credited in the existing accident analysis for the remaining applicable postulated accidents; therefore, they do not contribute to the margin of safety associated with the accident analysis. Certain postulated DBAs [design-basis accidents] involving the reactor are no longer possible because the reactor will be permanently shut down and defueled and QCNPS will no longer be authorized to operate the reactor.

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

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

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

Acting NRC Branch Chief: G. Edward Miller.

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

Date of amendment request: April 4, 2016, as supplemented by letters dated September 1, 2016, and November 10, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML16110A266, ML16260A399, and ML16323A313, respectively.

Description of amendment request: The amendments would revise the Technical Specification (TS) requirements for snubbers and add a new TS to the Administrative Controls section of the TSs describing the licensee's Snubber Testing Program. The amendments would revise the snubber TS surveillance requirements (SRs) by deleting specific requirements from the TS SRs and replacing them with a requirement to demonstrate snubber operability in accordance with the licensee-controlled Snubber Testing Program. The proposed changes include additions to, deletions from, and conforming administrative changes to the TSs.

The license amendment request was originally noticed in the Federal Register (FR) on July 5, 2016 (81 FR 43652). The notice is being reissued in its entirety because the licensee's supplement dated November 10, 2016, expanded the scope of the application by proposing to delete a portion of the snubber SR that requires inspections per another TS that is no longer applicable to snubbers.

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

Response: No.

The proposed changes would revise TS SR 4.7.6 to conform the TS to the revised surveillance program for snubbers. Snubber examination, testing and service life monitoring will continue to meet the requirements of 10 CFR 50.55a(g).

Snubber examination, testing and service life monitoring is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased.

Snubbers will continue to be demonstrated OPERABLE by performance of a program for examination, testing and service life monitoring in compliance with 10 CFR 50.55a or authorized alternatives. The proposed change to the TS 3.7.6 Action for inoperable snubbers is administrative in nature and is required for consistency with the proposed change to TS SR 4.7.6. The proposed change does not adversely affect plant operations, design functions or analyses that verify the capability of systems, structures, and components to perform their design functions[;] therefore, the consequences of accidents previously evaluated are not significantly increased.

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 changes do not involve any physical alteration of plant equipment. The proposed changes do not alter the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

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

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

Response: No.

The proposed changes ensure snubber examination, testing and service life monitoring will continue to meet the requirements of 10 CFR 50.55a(g). Snubbers will continue to be demonstrated OPERABLE by performance of a program for examination, testing and service life monitoring in compliance with 10 CFR 50.55a or authorized alternatives.

The proposed change to the TS 3.7.6 Action for inoperable snubbers is administrative in nature and is required for consistency with the proposed change to TS SR 4.7.6.

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

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

Acting NRC Branch Chief: Jeanne A. Dion.

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

Date of amendment request: September 16, 2016. A publicly-available version is in ADAMS under Accession No. ML16271A181.

Description of amendment request: The amendments would revise the Unit 1 and Unit 2 Technical Specifications (TSs) by removing certain process radiation monitors and placing their requirements in a licensee-controlled manual. The amendments would also change the Unit 2 containment particulate radiation monitor range.

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

Response: No.

The effluent radiation monitors are not event initiators, nor are they credited in the mitigation of any event or credited in the PRA [Probabilistic Risk Assessment]. Relocating the monitors to the ODCM [Offsite Dose Calculation Manual] does not adversely impact the monitor function, and does not affect the accident analyses in any manner.

The Unit 2 containment atmosphere particulate radiation monitor is credited in the Leak-Before-Break analyses, where it states that “the leakage detection systems are capable of detecting the specified leak rate” and that the leakage detection systems “are consistent with Regulatory Guide 1.45.” Correcting the TS instrument range for the monitor does not adversely impact the monitor function, i.e., its capability to detect leakage. This change does not affect the accident analyses in any manner.

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 previously evaluated?

Response: No.

The proposed changes correct a legacy error in the Unit 2 TS, and the TS removal of effluent monitors and their subsequent relocation to the ODCM do not change the function or capabilities of any equipment, and do not involve the addition or modification of any plant equipment. Also, the proposed change does not alter the design, configuration, or method of operation of the plant. The subject monitors remain capable of performing their design functions.

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

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

Response: No.

The proposed changes remove select effluent monitors from the TSs and relocate their requirements to the ODCM and correct a legacy error in the Unit 2 TSs, and do not involve the addition or modification of any plant equipment. The changes do not modify the plant or plant equipment, and do not change the manner in which structures, systems or components are design[ed] or evaluated.

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

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

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

Acting NRC Branch Chief: Jeanne A. Dion.

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment request: October 18, 2016. A publicly-available version is in ADAMS under Accession No. ML16294A257.

Description of amendment request: The proposed changes would revise Technical Specification 5.5.14, “Containment Leakage Rate Testing Program,” to clarify the containment leak rate testing pressure criteria.

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

Response: No.

The proposed changes do not involve changes to the installed structures, systems or components of the facility. The proposed change is consistent with Westinghouse Owners Group Standard Technical Specification language for the Containment Leak Rate Program.

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 does not introduce a new mode of plant operation and does not involve physical modification to the plant. The change does not introduce new accident initiators or impact assumptions made in the safety analysis. Testing requirements continue to demonstrate that the Limiting Conditions for Operation are met and the system components are functional.

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 change does not exceed or alter a design basis or safety limit, so there is no significant reduction in the margin of safety.

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

Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: David J. Wrona.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2 (DCPP), San Luis Obispo County, California

Date of amendment request: October 25, 2016. A publicly-available version is in ADAMS under Accession No. ML16315A184.

Description of amendment request: The proposed amendments would revise the Emergency Plan (E-Plan) for DCPP to adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” November 2012 (ADAMS Accession No. ML12326A805). NEI-99-01, Revision 6, has been endorsed by the NRC by letter dated March 28, 2013 (ADAMS Accession No. ML12346A463). Currently approved E-Plan EAL schemes for DCPP are based on the guidance established in NEI 99-01, Revision 4, “Methodology for Development of Emergency Action Levels,” January 2003 (ADAMS Accession No. ML030230250), except for security-related EALs, which are based on the guidance established in NEI 99-01, Revision 5, “Methodology for Development of Emergency Action Levels,” February 2008 (ADAMS Accession No. ML080450149).

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

Response: No.

The proposed changes to the Diablo Canyon Power Plant (DCPP) emergency action levels (EALs) do not impact the physical function of plant structures, systems, or components (SSCs) or the manner in which SSCs perform their design function. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are affected by the proposed changes.

Therefore, the proposed 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 changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed or removed) or a change in the method of plant operation. The proposed changes will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. The proposed changes to the DCPP EALs are not initiators of any accidents.

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

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

Response: No.

The proposed changes do not impact operation of the plant or its response to transients or accidents. The proposed changes do not affect the Technical Specifications or the Operating License. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. The E-Plan will continue to activate an emergency response commensurate with the extent of degradation of plant safety.

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

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: October 11, 2016. A publicly-available version is in ADAMS under Accession No. ML16285A351.

Description of amendment request: The amendment would add Technical Specification (TS) requirements for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9, consistent with NRC-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-427, Revision 2, “Allowance for Non-Technical Specification Barrier Degradation on Supported System OPERABILITY.” The availability of this TS improvement was published in the Federal Register on October 3, 2006 (71 FR 58444), as part of the consolidated line item improvement process (CLIIP).

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 by confirming the applicability of the Farley, Units 1 and 2, to the model proposed no significant hazards consideration published on October 3, 2006, as part of the CLIIP, as referenced below:

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

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

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

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

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

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

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

NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation (WCNOC), Docket No. 50-482, Wolf Creek Generating Station (WCGS), Coffey County, Kansas

Date of amendment request: September 30, 2016. A publicly-available version is in ADAMS under Accession No. ML16279A377.

Description of amendment request: The amendment would revise the emergency action level (EAL) scheme used at WCGS. The currently approved EAL scheme is based on Nuclear Management and Resources Council/National Environmental Studies Project (NUMARC/NESP)-007, Revision 2, “Methodology for Development of Emergency Action Levels,” January 1992 (ADAMS Accession No. ML041120174). The proposed change would allow WCNOC to adopt an EAL scheme which is based on the guidance established in Nuclear Energy Institute (NEI) 99-01, Revision 6, “Development of Emergency Action levels for Non-Passive Reactors,” November 2012 (ADAMS Accession No. ML12326A805). NEI 99-01, Revision 6 has been endorsed by the NRC by letter dated March 28, 2013 (ADAMS Accession No. ML12346A463).

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

Response: No.

The proposed changes to the WCGS EALs do not impact the physical function of plant structures, systems or components [(SSCs)] or the manner in which SSCs perform their design function. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are affected by the proposed changes.

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

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

Response: No.

The proposed changes do not involve a physical alteration of the plant (i.e., no new or different types of equipment will be installed or removed) or a change in the method of plant operation. The proposed changes will not introduce failure modes that could result in a new accident, and the changes do not alter assumptions made in the safety analysis. The proposed changes to the WCGS EALs are not initiators of any accidents.

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

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

Response: No.

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

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

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW., 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 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.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

Date of amendment request: December 10, 2015, as supplemented by letters dated March 2, July 7, and October 6, 2016.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.1.3, “Moderator Temperature Coefficient (MTC),” and TS 5.6.5, “Core Operating Limits Report (COLR),” to allow exemption from the normally required near end-of-life MTC measurement by placing a set of conditions on reactor core operation. If these conditions are met, the MTC measurement could be replaced by a calculated value.

Date of issuance: November 15, 2016.

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

Amendment Nos.: Unit 2-285; Unit 3-261. A publicly-available version is in ADAMS under Accession No. ML16215A243; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-26 and DPR-64: The amendments revised the Facility Operating Licenses and the TSs.

Date of initial notice in Federal Register: April 5, 2016 (81 FR 19647). The supplemental letters dated March 2, July 7, and October 6, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 15, 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: December 3, 2015, as supplemented by letters dated June 9, 2016, August 2, 2016, and November 8, 2016.

Brief description of amendments: The amendments revised the technical specification (TS) surveillance requirements associated with the emergency diesel generator (EDG) fuel oil transfer system. Specifically, the amendments allow for the crediting of manual actions, in lieu of automatic actions, without having to declare the EDGs inoperable.

Date of issuance: November 16, 2016.

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

Amendments Nos.: Unit 2-311; Unit 3-315. A publicly-available version is in ADAMS under Accession No. ML16292A188; 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: February 2, 2016 (81 FR 5498). The supplemental letters dated June 9, 2016, August 2, 2016, and November 8, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 16, 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: November 20, 2015, as supplemented by letters dated January 12, April 11, and June 30, 2016.

Brief description of amendments: The amendments revised the setpoint requirements in Technical Specification (TS) 3.3.5, “Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.” The change was requested to fulfill a license condition to eliminate the manual actions in lieu of automatic degraded voltage protection to assure adequate voltage to safety-related equipment during design-basis events.

Date of issuance: November 17, 2016.

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

Amendment Nos.: Unit 1-206; Unit 2-202. A publicly-available version is in ADAMS under Accession No. ML16196A161; 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 16, 2016 (81 FR 7842). The supplemental letters dated April 11, 2016, and June 30, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 17, 2016.

No significant hazards consideration comments received: No.

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

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

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

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

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

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

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

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

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

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

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

B. Electronic Submissions (E-Filing)

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

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

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

Date of amendment request: September 30, 2016, as supplemented by letter dated November 8, 2016.

Description of amendment request: The amendments revised the LSCS licensing basis related to Alternate Source Term Analysis in the Updated Final Safety Analysis Report to allow operation with and movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 megawatt days per metric ton of uranium (MWD/MTU), which is the current rod average burnup limit specified in Footnote 11 of NRC Regulatory Guide (RG) 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” to which LSCS is committed. In addition, the change allows use of the release fractions listed in Table 1 of RG 1.183 for these Atrium-10 partial length rods that are currently in the LSCA, Unit 2, Cycle 16, reactor core for the remainder of the current operating cycle.

Date of issuance: November 18, 2016.

Effective date: As of the date of issuance and shall be implemented within 10 days from the date of issuance.

Amendment Nos.: Unit 1—221; Unit 2—207. A publicly-available version is in ADAMS under Accession No. ML16320A182; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the licensing basis related to Alternate Source Term in the Updated Final Safety Analysis Report.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Public notice of the proposed amendment was published in The Ottawa Times on November 15 and November 16, 2016. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated November 18, 2016.

Attorney for licensee: Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Acting Branch Chief: Edward G. Miller.

Dated at Rockville, Maryland, this 23rd day of November 2016.

For the Nuclear Regulatory Commission.

Anne T. Boland,

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

[FR Doc. 2016-28990 Filed 12-5-16; 8:45 am]

BILLING CODE 7590-01-P