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Notice

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

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Start Preamble

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 October 2, 2014 to October 15, 2014. The last biweekly notice was published on October 14, 2014.

DATES:

Comments must be filed by November 28, 2014. A request for a hearing must be filed by December 29, 2014.

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-2014-0239. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-3422; 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: 3WFN-06-A44M, 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.

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FOR FURTHER INFORMATION CONTACT:

Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov

End Further Info End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

Start Printed Page 64220

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2014-0239 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-2014-0239.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section.
  • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2014-0239 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

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 should 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. Should the Commission take 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. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

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

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

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

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of Start Printed Page 64221which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, 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.

B. Electronic Submissions (E-Filing)

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

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

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

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

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

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

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having Start Printed Page 64222granted 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, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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

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

Date of amendment request: October 29, 2013, as supplemented by letters dated May 7, 2014 and June 17, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML13316C083, ML14139A006, and ML14178B284.

Description of amendment request: The amendment would revise the CR-3 Facility Operating License (FOL) to remove and revise certain License Conditions. This amendment also proposes to extensively revise the CR-3 Improved Technical Specifications (ITS) in order to create the CR-3 Permanently Defueled Technical Specifications (PDTS).

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

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

Response: No.

CR-3 has permanently ceased operation. The proposed amendment would modify the CR-3 FOL and ITS by proposing to delete certain License Conditions (LCs) and ITS that are no longer applicable to a permanently defueled facility, while modifying the remaining portions to correspond to the permanently shutdown condition. Changes proposed to LCs will make them consistent with the non-operating status of CR-3. Other proposed LCs changes will eliminate LCs that were designed for one time implementation and have been satisfied, or are no longer required due to changes to Part 50 or Part 73 regulations that accomplish the same result or eliminate the requirement for the LC. The proposed changes to the ITS are consistent with the criteria set forth in 10 CFR 50.36 for the contents of ITS.

Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR) described the design basis accident (DBA) and transient scenarios applicable to CR-3 during power operations. With the reactor in a permanently defueled condition, the spent fuel pool and its cooling systems are dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents is much smaller than for an operational plant. As a result of the certifications submitted by CR-3 in accordance with 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactor or to place or retain fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2), the majority of the accident scenarios originally postulated in the FSAR are no longer possible and have been removed from the FSAR under 10 CFR 50.59.

The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant boundary;

2. The capability to shutdown the reactor and maintain it in a safe shutdown condition; or

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1) or 100.11.

The first two criteria, integrity of the reactor coolant pressure boundary and safe shutdown of the reactor, are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after the termination of reactor operations at CR-3 and the permanent removal of the fuel from the reactor vessel (following 4 years of decay time after shutdown) and purging of the contents of the waste gas decay tanks, none of the SSCs at CR-3 are required to be relied on for accident mitigation. Therefore, none of the SSCs at CR-3 meet the definition of a safety-related SSC stated in 10 CFR 50.2 (with the exception of the passive spent fuel pool structure).

The deletion of ITS definitions and rules of usage and application, that are currently not applicable in a defueled condition, has no impact on facility SSCs or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shutdown and defueled status of CR-3 has no impact on the remaining DBA [design basis accidents] (the Fuel Handling Accident in the Auxiliary Building) or the proposed Radioactive Waste Handling Accident. The removal of LCOs [Limiting Conditions for Operation] or SRs [Surveillance Requirements] that are related only to the operation of the nuclear reactor or accidents do not affect mitigation of the applicable DBAs previously evaluated since these DBAs are no longer applicable in the defueled mode. The safety functions involving core reactivity control, reactor heat removal, reactor coolant system inventory control, and containment integrity are no longer applicable at CR-3 as a permanently defueled plant. The analyzed accidents involving damage to the reactor coolant system, main steam lines, reactor core, and the subsequent release of radioactive material are no longer possible at CR-3

Since CR-3 has permanently ceased operation, the generation of fission products has ceased and the remaining source term will decay. The radioactive decay of the irradiated fuel since shutdown of the reactor have reduced the consequences of the Fuel Handling Accident (FHA) to levels well below those previously analyzed. The relevant parameter (water level) associated with the fuel pool provides an initial condition for the FHA analysis and is included in the PDTS.

The spent fuel pool water level, spent fuel pool boron concentration, and spent fuel pool storage LCOs are retained to preserve the current requirements for safe storage of irradiated fuel.

Fuel pool cooling and makeup related equipment and support equipment (e.g., electrical power systems) are not required to be continuously available since there is sufficient time to effect repairs, establish alternate sources of makeup flow, or establish alternate sources of cooling in the event of a loss of cooling and makeup flow to the spent fuel pool.

The deletion and modification of provisions of the Administrative Controls do not directly affect the design of SSCs necessary for the safe storage of irradiated fuel or the methods used for handling and Start Printed Page 64223storage of such fuel in the fuel pool. Deletion of Programs are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shutdown and defueled condition of the reactor.

The proposed LC revisions reflect the CR-3 functions that are still authorized in the permanently defueled condition, and remove authorizations that suggest the reactor can be placed in operation. LCs that are being removed due to their one time applicability being previously satisfied have no bearing on future functions at CR-3. Other LCs are being removed that are not required by regulation for a permanently defueled and decommissioning plant. These changes cannot increase the probability or consequences of any accident that remains credible. The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition is the only operation currently allowed, and is therefore bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

Therefore, the proposed amendment 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 have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The removal of ITS that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor is permanently shutdown and defueled, and CR-3 is no longer authorized to operate the reactor.

The proposed deletion of requirements of the CR-3 ITS do not affect safe storage of nuclear fuel. The proposed PDTS continue to require proper control and monitoring of safety significant parameters. The proposed restriction on the fuel pool level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA. The spent fuel pool water level, spent fuel pool boron concentration, and spent fuel pool storage LCOs are retained to preserve the current requirements for safe storage of irradiated fuel.

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

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

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

Response: No.

Because the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation are no longer credible. The only remaining credible accident is a FHA. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact a FHA.

The proposed changes are limited to those portions of the LCs and ITS that are not related to the safe storage of irradiated fuel. The requirements for SSCs that have been deleted from the CR-3 ITS are not credited in the existing accident analysis for the remaining applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated DBAs involving the reactor are no longer possible because the reactor is permanently shutdown and defueled and CR-3 is no longer authorized to operate the reactor.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety because the current design limits continue to be met for the accident of concern.

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

NRC Branch Chief: Douglas A. Broaddus.

Entergy Nuclear Indian Point 3, LLC, Docket No. 50-286, Indian Point, Unit 3, Westchester County, NY

Date of amendment request: April 1, 2014. A publicly-available version is in ADAMS under Accession No. ML14099A333.

Description of amendment request: The amendment would revise Technical Specifications (TS) Figure 3.4.3-1, Heatup Limitations for Reactor Coolant System, Figure 3.4.3-2, Cooldown Limitations for Reactor Coolant System, and Figure 3.4.3-3, Hydrostatic and Inservice Leak Testing Limitations for Reactor Coolant System, to indicate that the curves are applicable for vacuum fill.

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

1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

The proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. There are no physical changes to the plant being introduced by the proposed changes to the heatup, cooldown and hydrostatic inservice leak testing limitation curves. The proposed changes do not modify the RCS pressure boundary. That is, there are no changes in operating pressure, materials, or seismic loading. The proposed changes do not adversely affect the integrity of the RCS pressure boundary such that its function in the control of radiological consequences is affected. The heatup, cooldown and hydrostatic inservice leak testing limitation curves were established in compliance with the methodology used to calculate and predict effects of radiation on embrittlement of RPV beltline materials and remain valid during vacuum fill.

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

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents.

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

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in the margin of safety.

The proposed TS changes do not involve a significant reduction in the margin of safety. The changes clarify that the heatup, cooldown and hydrostatic inservice leak testing limitation curves remain valid during vacuum fill (to 0 psia) in accordance with current regulations. Because operation will be within these limits, the RCS materials will continue to behave in a non-brittle manner consistent with the original design bases.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff Start Printed Page 64224proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Benjamin Beasley.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois

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

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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

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

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

Date of amendment request: July 14, 2014. A publicly-available version is in ADAMS under Accession No. ML14195A172.

Description of amendment request: The proposed amendment would revise and add technical specification (TS) surveillance requirements to address the concerns discussed in NRC Generic Letter 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” dated January 11, 2008 (ADAMS Accession No. ML072910759). The proposed TS changes are based on NRC-approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation,” dated February 21, 2013 (ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of Availability for TSTF-523, Revision 2, for plant-specific adoption using the Consolidated Line Item Improvement Process, in the Federal Register on January 15, 2014 (79 FR 2700).

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

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

Response: No.

The proposed change revises or adds Surveillance Requirements (SRs) that require verification that the Emergency Core Cooling System (ECCS), the Decay Heat Removal (DHR)/Residual Heat Removal (RHR)/Shutdown Cooling (SDC) System, the Containment Spray (CS) System, and the Reactor Core Isolation Cooling (RCIC) System, as applicable, are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable to perform their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased.

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

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 revises or adds SRs that require verification that the ECCS, the DHR/RHR/SDC System, the CS System, and the RCIC System, as applicable, are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions.

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

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

Response: No.

The proposed change revises or adds SRs that require verification that the ECCS, the DHR/RHR/SDC System, the CS System, and the RCIC System, as applicable, are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.

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 requested amendments involve no significant hazards consideration.

Attorney for licensee: J. Bradley Fewell, Esquire, Vice President and Deputy General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Travis L. Tate.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station (BVPS), Units 1 and 2, Beaver County, Pennsylvania

Date of amendment request: September 4, 2014. A publicly-available version is available in ADAMS under Accession No. ML14247A512.

Description of amendment request: The amendment proposes changes to align the BVPS Emergency Planning Zone (EPZ) boundary with the boundary that is currently in use by the emergency management agencies of the three counties that implement public protective actions around BVPS.

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

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

Response: No.

This amendment request would alter portions of the outer EPZ boundary defined in the BVPS EPP [Emergency Preparedness Plan] to align with the EPZ boundaries implemented by the Columbiana County, Hancock County, and Beaver County emergency management agencies. The proposed amendment does not involve any modifications or physical changes to plant systems, structures, or components. The proposed amendment does not change plant operations or maintenance of plant systems, structures, or components. Nor does the proposed amendment alter any BVPS EPP Start Printed Page 64225facility or equipment. Changing the EPZ boundaries cannot increase the probability of an accident since emergency plan functions would be implemented after a postulated accident occurs. The proposed amendment does not alter or prevent the ability of the BVPS emergency response organization to perform intended emergency plan functions to mitigate the consequences of and to respond adequately to radiological emergencies.

Therefore, the proposed amendment 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.

This amendment request alters the EPZ boundary described in the BVPS EPP. The proposed amendment does not involve any design modifications or physical changes to the plant, does not change plant operation or maintenance of equipment, and does not alter BVPS EPP facilities or equipment. The proposed amendment to the BVPS EPP does not alter any BVPS emergency actions that would be implemented in response to postulated accident events. The proposed amendment does not create any credible new failure mechanisms, malfunctions, or accident initiators not previously considered.

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.

This amendment request would alter portions of the EPZ boundary defined in the BVPS EPP. The proposed amendment does not involve any design or licensing basis functions of the plant, no physical changes to the plant are made, does not impact plant operation or maintenance of equipment, and does not alter BVPS EPP facilities or equipment. This change does not alter any BVPS emergency actions that would be implemented in response to postulated accident events. The BVPS EPP continues to meet 10 CFR 50.47 and 10 CFR 50, Appendix E requirements for emergency response.

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: David W. Jenkins, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Meena K. Khanna.

Florida Power and Light Company, et al. (the licensee), Docket Nos. 50-335 and 50-389, St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida

Date of amendment request: July 14, 2014. A publicly-available version is in ADAMS under Accession No. ML14198A074.

Description of amendment request: The amendments would revise or add surveillance requirements (SRs) to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances that permit performance of the verification. The licensee proposed the changes to address NRC Generic Letter 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems” (ADAMS Accession No. ML072910759), as described in Revision 2 of Technical Specification Task Force No. 523, “Generic Letter 2008-01, Managing Gas Accumulation” (ADAMS Accession No. ML13053A075).

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

1. Does the Proposed Change Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated?

Response: No.

The proposed change revises or adds SRs that require verification that the Emergency Core Cooling Systems (ECCS), Residual Heat Removal (RHR) System, Shutdown Cooling (SDC) System, and Containment Spray (CS) System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable of performing their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased.

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

2. Does the Proposed Change Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated?

Response: No.

The proposed change revises or adds SRs that require verification that the ECCS, RHR System, SDC System, and CS System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions.

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

3. Does the Proposed Change Involve a Significant Reduction in a Margin of Safety?

Response: No.

The proposed change revises or adds SRs that require verification that the ECCS, RHR System, SDC System, and CS System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure that the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.

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, 700 Universe Blvd., MS LAW/JB, Juno Beach, Florida 33408-0420.

NRC Acting Branch Chief: Lisa M. Regner.

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

Date of amendment request: August 8, 2014. A publicly-available version is in ADAMS under Accession No. ML14225A654.

Description of amendment request: The amendments would modify the Technical Specifications by removing TS 3/4.4.7, “Chemistry,” which provides limits on the oxygen, chloride, and fluoride content in the reactor coolant system to minimize corrosion. Start Printed Page 64226The licensee requested that these requirements be relocated to the Updated Final Safety Analysis Report (UFSAR) and controlled in accordance with 10 CFR 50.59, “Changes, tests, and experiments.”

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

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

Response: No.

The proposed change acts to remove current Reactor Coolant System (RCS) chemistry limits and monitoring requirements from the TS and relocate the requirements to the UFSAR. Monitoring and maintaining RCS chemistry minimizes the potential for corrosion of RCS piping and components. Corrosion effects are considered a long-term impact on RCS structural integrity. Because RCS chemistry will continue to be monitored and controlled, relocating the current TS requirements to the UFSAR will not present an adverse impact to the RCS and subsequently, will not impact the probability or consequences of an accident previously evaluated. Furthermore, once relocated to the UFSAR, changes to RCS chemistry limits and monitoring requirements will be controlled in accordance with 10 CFR 50.59.

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

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

Response: No.

The proposed change acts to remove current Reactor Coolant System (RCS) chemistry limits and monitoring requirements from the TS and relocate the requirements to the UFSAR. The proposed change does not introduce new modes of plant operation and it does not involve physical modifications to the plant (no new or different type of equipment will be installed). There are no changes in the method by which any safety related plant structure, system, or component (SSC) performs its specified safety function. As such, the plant conditions for which the design basis accident analyses were performed remain valid.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of the proposed change. There will be no adverse effect or challenges imposed on any SSC as a result of the proposed change.

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

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

Response: No.

Margin of safety is related to confidence in the ability of the fission product barriers to perform their accident mitigation functions. The proposed change acts to remove current Reactor Coolant System (RCS) chemistry limits and monitoring requirements from the TS and relocate the requirements to the UFSAR. The proposed change will maintain limits on RCS chemistry parameters and will continue to provide associated monitoring requirements. The proposed change does not physically alter any SSC. There will be no effect on those SSCs necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, loss of cooling accident peak cladding temperature (LOCA PCT), or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met.

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

The NRC staff 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 Blvd. MS LAW/JB, Juno Beach, Florida 33408-0420.

NRC Acting Branch Chief: Lisa M. Regner.

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

Date of amendment request: July 1, 2014. A publicly-available version is in ADAMS under Accession No. ML14192A338.

Description of amendment request. The amendments would revise Technical Specification (TS) 3.8.1, “AC [Alternating Current] Sources—Operating,” to extend, on a one-time basis, the Completion Time (CT) of Required Action A.3 from 72 hours to 14 days. By letter dated September 18, 2013 (ADAMS Accession No. ML13232A143), the NRC staff issued Amendment No. 160 to Facility Operating License No. NPF-87 and Amendment No. 160 to Facility Operating License No. NPF-89 for CPNPP, Units 1 and 2, respectively. The amendments revised TS 3.8.1 to extend the CT for Required Action A.3 on a one-time basis from 72 hours to 14 days. The CT extension from 72 hours to 14 days was to be used twice while completing the plant modification to install alternate startup transformer (ST) XST1A and was to expire on March 31, 2014.

The first 14-day CT was successfully completed, on October 14, 2013. However, the licensee inadvertently cut the wrong offsite power cable during the second 14-day CT resulting in a total loss-of-offsite power (LOOP) to both units and the modification had to be abandoned. Due to the cut-cable event and the subsequent efforts to determine the causes and corrective actions, the modification could not be completed by March 31, 2014. The licensee has requested an extension of the CT from 72 hours to 14 days on one-time basis to complete the plant modification. Installation of the alternate ST XST1A will result in improved offsite power system reliability.

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

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

Response: No.

The proposed change will revise the CT for the loss of one offsite source from 72 hours to 14 days to allow a one-time, 14-day CT. The proposed one-time extension of the CT for the loss of one offsite power circuit does not significantly increase the probability of an accident previously evaluated. The TS will continue to require equipment that will power safety related equipment necessary to perform any required safety function. The one-time extension of the CT to 14 days does not affect the design of the STs, the interface of the STs with other plant systems, the operating characteristic of the STs, or the reliability of the STs.

The consequence of a LOOP event has been evaluated in the CPNPP Final Safety Analysis Report (Reference 8.1 [of the licensee's letter dated July 1, 2014]) and the Station Blackout evaluation. Increasing the CT for one offsite power source on a one-time basis from 72 hours to 14 days does not increase the consequences of a LOOP event nor change the evaluation of LOOP events.

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

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

Response: No.

The proposed change does not result in a change in the manner in which the electrical distribution subsystems provide plant protection. The proposed change will only affect the time allowed to restore the operability of the offsite power source through a ST. The proposed change does not affect the configuration, or operation of the Start Printed Page 64227plant. The proposed change to the CT will facilitate installation of a plant modification which will improve plant design and will eliminate the necessity to shut down both Units if XST1 fails or requires maintenance that goes beyond the current TS CT of 72 hours. This change will improve the long-term reliability of the 138 [kiloVolt (kV)] offsite circuit ST which is common to both CPNPP Units.

There are no changes to the STs or the supporting systems operating characteristics or conditions. The change to the CT does not change any existing accident scenarios, nor create any new or different accident scenarios. In addition, the change does not impose any new or different requirements or eliminate any existing requirements. The change does not alter any of the assumptions made in the safety analysis.

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

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

Response: No.

The proposed change does not affect the acceptance criteria for any analyzed event nor is there a change to any safety limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. Neither the safety analyses nor the safety analysis acceptance criteria are affected by this change. The proposed change will not result in plant operation in a configuration outside the current design basis. The proposed activity only increases a one-time pre-planned occurrence, the period when the plant may operate with one offsite power source. The margin of safety is maintained by maintaining the ability to safely shut down the plant and remove residual heat.

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

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

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

NRC Branch Chief: Michael T. Markley.

Northern States Power Company—Minnesota, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: August 21, 2014. A publicly-available version is in ADAMS under Accession No. ML14233A431.

Description of amendment request: The proposed amendments would revise the PINGP, Units 1 and 2, licensing basis analysis for waste gas decay tank rupture.

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

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

Response: No.

This license amendment request proposes to revise the licensing basis waste gas decay tank rupture analysis. The proposed analysis was updated to include the current fuel type, current fuel cycle lengths and plant operation to sixty years.

The proposed waste gas decay tank rupture analysis changes are not accident initiators, and therefore the proposed changes do not involve an increase in the probability of an accident.

The original waste gas decay tank rupture analysis demonstrated that the doses were a small fraction of the regulatory guidelines and that the waste gas system design prevents release of undue amounts of radioactivity. The revised waste gas decay tank rupture analysis demonstrates that the doses are well within the regulatory guidelines and that the waste gas system design continues to prevent release of undue amounts of radioactivity, and thus the proposed changes do not involve a significant increase in the consequences of an accident.

Therefore, the proposed licensing basis 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.

This license amendment request proposes to revise the licensing basis waste gas decay tank rupture analysis. The proposed analysis was updated to include the current fuel type, current fuel cycle lengths and plant operation to sixty years.

The proposed waste gas decay tank rupture analysis includes plant changes that have previously been evaluated. This analysis applies the same methodology as the previous analysis. The proposed revision to the waste gas decay tank rupture analysis does not change any system operations or maintenance activities. The changes do not involve physical alteration of the plant; that is, no new or different type of equipment will be installed. These changes do not create new failure modes or mechanisms which are not identifiable during testing and no new accident precursors are generated.

Therefore, the proposed licensing basis 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.

This license amendment request proposes to revise the licensing basis waste gas decay tank rupture analysis. The proposed analysis was updated to include the current fuel type, current fuel cycle lengths and plant operation to sixty years.

This revised analysis applies the same methodology as the original waste gas decay tank rupture analysis. The original waste gas decay tank rupture analysis demonstrated that the doses were a small fraction of the regulatory guidelines and that the waste gas system design prevents release of undue amounts of radioactivity. The revised waste gas decay tank rupture analysis demonstrates that the doses are well within the regulatory guidelines and that the waste gas system design continues to prevent release of undue amounts of radioactivity.

Therefore, the proposed licensing basis 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 requests involve no significant hazards consideration.

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401

NRC Branch Chief: David L. Pelton.

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

Date of amendment request: July 28, 2014. A publicly-available version is in ADAMS under Accession No. ML14210A484.

Description of amendment request: The proposed amendment would modify technical specification (TS) requirements regarding steam generator tube inspections and reporting as described in Technical Specification Task Force (TSTF) traveler TSTF-510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.” In addition, the proposed amendment would revise the SNGS, Unit 2, TSs 6.8.4.i, “Steam Generator (SG) Program,” TS 6.9.1.10, “Steam Generator Tube Inspection Report,” and the bases section of 3/4.4.6, “Steam Generator (SG) Tube Integrity,” to remove unnecessary information related to the original Salem Unit 2 Westinghouse steam generators.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented Start Printed Page 64228below, along with NRC edits in square brackets:

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

Response: No.

The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to the Salem Unit 2 Technical Specifications (TS) that are not associated with TSTF-510, removing unnecessary information related to W* [pronounced “W star,” which refers to the length of the steam generator tube required to be inspected within the hot-leg tube sheet] that is only applicable to Westinghouse steam generators, is an administrative change that 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 to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

The proposed changes to the Salem Unit 2 Technical Specifications (TS) that are not associated with TSTF-510, removing unnecessary information related to W* that is only applicable to Westinghouse steam generators, is an administrative change that does not affect the design of the SGs or their method of operation.

Therefore, it is concluded that these changes do not create the possibility of a new of different kind of accident from any accident previously evaluated.

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

Response: No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

The proposed changes to the Salem Unit 2 Technical Specifications (TS) that are not associated with TSTF-510, removing unnecessary information related to W* that is only applicable to Westinghouse steam generators, is an administrative change that does not involve a significant reduction in a margin of safety.

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

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

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

NRC Branch Chief: Meena K. Khanna.

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

Date of amendment request: June 20, 2014, as supplemented by letter dated August 6, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14174B176 and ML14218A809.

Description of amendment request: The proposed license amendment request (LAR) would revise the Updated Final Safety Analysis Report (UFSAR) in regard to Tier 2* information related to fire area boundaries. These changes add three new fire zones in the middle annulus to provide enclosures for the Class 1E electrical containment penetrations in accordance with UFSAR Appendix 9A, Subsection 9A.3.1.1.15. The addition of the three new fire zones extended the fire area boundaries for three existing fire areas and, therefore, constitutes a change to Tier 2* information. Additionally, the licensee proposed changes that require revisions to UFSAR Tier 2 information involving changes to plant-specific Tier 2* information.

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

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

Response: No.

The proposed middle annulus fire barrier reconfiguration for the electrical penetrations would not adversely affect any safety-related equipment or function. The modified configuration for the Class 1E electrical containment penetration enclosures will maintain the fire protection function (i.e., barrier) as evaluated in Updated Final Safety Analysis Report (UFSAR), thus, the probability of a Class 1E electrical containment penetration failure is not significantly increased. The safe shutdown fire analysis is not affected, and the fire protection analysis results are not adversely affected. The proposed changes do not involve any accident, initiating event or component failure; thus, the probabilities of previously evaluated accidents are not affected. The maximum allowable leakage rate specified in the Technical Specifications is unchanged, and radiological material release source terms are not affected; thus, the radiological releases in the accident analyses are not affected.

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

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

Response: No.

The addition of enclosures constructed of three-hour rated fire barriers to separate the fire zones in the middle annulus for the Class 1E electrical penetration assemblies will maintain the fire protection function as evaluated in the UFSAR. The addition of the fire barriers does not affect the function of the Class 1E electrical containment penetrations or electrical penetration assemblies, and thus, does not introduce a new failure mode. The addition of the fire barriers does not create a new fault or sequence of events that could result in a radioactive material release.

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

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

Response: No.

The use of enclosures constructed of three-hour rated fire barriers to separate the fire zones in the middle annulus for the Class 1E electrical penetration assemblies will maintain the fire protection function as evaluated in the UFSAR. The use of the fire barriers does not affect the ability of the Class Start Printed Page 642291E electrical containment penetrations, electrical penetration assemblies, or the containment to perform their design function. The Class 1E electrical containment penetrations and electrical penetration assemblies within the enclosures continue to comply with the existing design codes and regulatory criteria, and do not affect any safety limit. The use of fire barriers and enclosures to separate the Class 1E electrical penetration assemblies does not adversely affect any margin of safety.

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

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

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

NRC Branch Chief: Lawrence J. Burkhart.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project (STP), Units 1 and 2, Matagorda County, Texas

Date of amendment request: May 15, 2014, as supplemented by letter dated July 10, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14164A341 and ML14282A185, respectively.

Description of amendment request: The amendment would update the Emergency Action Levels (EALs) used at STP, Units 1 and 2 from the current scheme based on Nuclear Management and Resources Council, Inc. (NUMARC)/Nuclear Environmental Studies Project (NESP) report NUMARC/NESP-007, Revision 2, “Methodology for Development of Emergency Action Levels,” dated January 1992 (ADAMS Accession No. ML041120174), to the NRC-endorsed scheme contained in Nuclear Energy Institute (NEI) 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” dated November 2012 (ADAMS Accession No. ML12326A805). The EAL scheme in NEI 99-01, Revision 6 includes an EAL for Independent Spent Fuel Storage Installations (ISFSI), which is needed in order to implement dry cask storage operations at STP Units 1 and 2. Additionally, there are three EALs that require Spent Fuel Pool level instrument values which are designed to address lessons learned from the Fukushima Dai-ichi accident.

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

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

Response: No.

The change revises the STPNOC [STP Nuclear Operating Company] Emergency Action Levels to be consistent with the NRC endorsed EAL scheme contained in NEI 99-01, Revision 6, “Methodology for Development of Emergency Action Levels,” but does not alter any of the requirements of the Operating License or the Technical Specifications. In addition to replacing the current STP EALs, the new EAL scheme includes an EAL related to the planned STP Independent Spent Fuel Storage Installation, and EALs related to planned changes to the Spent Fuel Pool level instrumentation that will address lessons learned from Fukushima Daiichi. The proposed change does not modify any plant equipment and does not impact any failure modes that could lead to an accident. Additionally, the proposed change has no effect on the consequences of any analyzed accident since the change does not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase 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 change revises the STPNOC Emergency Action Levels to be consistent with the NRC endorsed EAL scheme contained in NEI 99-01, Revision 6, “Methodology for Development of Emergency Action Levels,” but does not alter any of the requirements of the Operating License or the Technical Specifications. In addition to replacing the current STP EALs, the new EAL scheme includes an EAL related to the planned STP Independent Spent Fuel Storage Installation, and EALs related to planned changes to the Spent Fuel Pool level instrumentation that will address lessons learned from Fukushima Daiichi. The proposed change does not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No system setpoints are being modified. No new failure modes are introduced by the proposed change. The proposed amendment does not introduce any accident initiators or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The change revises the STPNOC Emergency Action Levels to be consistent with the NRC endorsed EAL scheme contained in NEI 99-01, Revision 6, “Methodology for Development of Emergency Action Levels,” but does not alter any of the requirements of the Operating License or the Technical Specifications. In addition to replacing the current STP EALs, the new EAL scheme includes an EAL related to the planned STP Independent Spent Fuel Storage Installation, and EALs related to planned changes to the Spent Fuel Pool level instrumentation that will address lessons learned from Fukushima Daiichi. The proposed change does not affect any of the assumptions used in the accident analysis, nor does it affect any operability requirements for equipment important to plant safety. Therefore, the proposed change will not result in a significant reduction in the margin of safety in operation of the facility as discussed in this license amendment request.

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

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.

NRC Branch Chief: Michael T. Markley.

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

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

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

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental Start Printed Page 64230impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.

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

Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam Electric Plant (HBRSEP), Unit 2, Darlington County, South Carolina

Date of application for amendment: September 30, 2013, as supplemented by letter dated August 6, 2014.

Brief description of amendment: The amendment implements the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-491, “Removal of Main Steam and Main Feedwater Valve Isolation Times from Technical Specifications,” via the Consolidated Line Item Improvement Process. This amendment modifies the current Technical Specifications (TSs) 3.7.2, Main Steam Isolation Valves and 3.7.3, Main Feedwater Isolation Valves, Main Feedwater Regulation Valves and Bypass Valves by relocating the specific isolation time for the isolation valves from the associated Surveillance Requirements (SRs). The isolation time in the TS SRs is replaced with the requirement to verify the valve isolation time is “within limits.” The specific isolation times will be maintained in the HBRSEP, Unit 2, Technical Requirements Manual.

Date of issuance: October 10, 2014.

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

Amendment No.: 237. A publicly-available version is in ADAMS under Accession No. ML14252A221; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-23. Amendment revised the Facility Operating License and TSs.

Date of initial notice in Federal Register : December 10, 2013 (78 FR 74180). The supplemental letter dated August 6, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 10, 2014.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: May 1, 2014, as supplemented by letter dated August 21, 2014.

Brief description of amendment: The amendment revises Technical Specification 2.0, “Safety Limits (SLs),” by changing the safety limit minimum critical power ratio for both single and dual recirculation loop operation.

Date of issuance: September 30, 2014.

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

Amendment No.: 307. A publicly-available version is in ADAMS under Accession No. ML14258B201; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-59: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register : August 5, 2014 (79 FR 45487). The supplemental letter dated August 21, 2014, 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 (NSHC) determination as published in the Federal Register.

The Commission's related evaluation of the amendment and final NSHC determination is contained in a Safety Evaluation dated September 30, 2014.

No significant hazards consideration comments received. No.

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 20, 2013, as supplemented by letter dated June 30, 2014.

Brief description of amendment: The amendment revised the LSCS, Units 1 and 2, allowable values for the loss of voltage relay voltage setpoints in Technical Specification Table 3.3.8.1-1, “Loss of Power Instrumentation.”

Date of issuance: September 29, 2014.

Effective date: As of the date of issuance. For LSCS Unit 1, the amendment shall be implemented prior to entering MODE 4 following the spring 2016 refueling outage (L1R16). For LSCS, Unit 2, the amendment shall be implemented prior to entering MODE 4 following the spring 2015 refueling outage (L2R15).

Amendment No.: 209 and 196. A publicly-available version is in ADAMS under Accession No. ML14252A913; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License Nos. NPF-11 and NPF-18: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : December 10, 2013 (78 FR 74182). The supplemental letter dated June 30, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2014.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit (TMI-1), Dauphin County, Pennsylvania

Date of application for amendment: May 7, 2014.

Brief description of amendment: The amendment revises Technical Specification (TS) Surveillance Requirements (SRs) 4.12.1, “Emergency Control Room Air Treatment System,” and 4.12.4, “Fuel Handling Building [Engineered Safety Feature] ESF Air Treatment System.” The amendment revised the TSs to replace the existing SRs to operate ventilation systems with charcoal filters for a 10-hour period at a frequency controlled in accordance with the Surveillance Frequency Control Program (SFCP) with a requirement to operate the systems for greater than or equal to 15 continuous minutes at a frequency controlled in accordance with the SFCP. These changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-522, Revision 0, “Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month.”

Date of issuance: October 14, 2014.Start Printed Page 64231

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

Amendment No.: 282. A publicly-available version is in ADAMS under Accession No. ML14240A348; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-50: Amendment revised the license and the technical specifications.

Date of initial notice in Federal Register: August 5, 2014 (79 FR 45476).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 14, 2014.

No significant hazards consideration comments received: No.

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

Date of amendment requests: April 9, 2014, as supplemented by letter dated August 15, 2014.

Brief description of amendments: The amendments revised the CNP Technical Specifications (TSs) 3.4.3, “[Reactor Coolant System] RCS Pressure and Temperature Limits.” The changes to TSs clarify that pressure limits are considered to be met for pressures that are below 0 psig (i.e., up to and including full vacuum conditions). Vacuum fill operations for the RCS can result in system pressures below 0 psig.

Date of issuance: October 1, 2014.

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

Amendment Nos.: 323 (Unit 1) and 306 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML14259A549; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments revise the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: July 8, 2014 (79 FR 38591). The supplemental letter dated August 15, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 1, 2014.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 18, 2014, as supplemented by the letter dated July 30, 2014.

Brief description of amendment: The license amendment revises the Updated Final Safety Analysis Report (UFSAR) in regard to Tier 2* information related to fire area boundaries. These changes add three new fire zones in the middle annulus to provide enclosures for the Class 1E electrical containment penetrations in accordance with UFSAR Appendix 9A, Subsection 9A.3.1.1.15. Additionally, the license amendment revises UFSAR Tier 2 information involving changes to plant-specific Tier 2* information.

Date of issuance: October 8, 2014.

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

Amendment No.: 25. A publicly-available version is in ADAMS under Accession No. ML14248A243; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: May 27, 2014 (79 FR 30189).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 8, 2014.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 27, 2014, as supplemented by the letter dated July 23, 2014.

Brief description of amendment: The amendment revises the VEGP Units 3 and 4 Emergency Plan and changes the combined licenses (COL), Appendix C, plant-specific emergency planning inspections, tests, analyses, and acceptance criteria (ITAAC) to reflect the relocation of the Operations Support Centers and changes the description of the plant monitoring system.

Date of issuance: October 7, 2014

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

Amendment No.: 24. A publicly-available version is in ADAMS under Accession No. ML14245A075; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: May 13, 2014 (79 FR 27345).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 7, 2014.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 29, 2014. A redacted version was provided by letter dated May 27, 2014.

Brief description of amendment: The amendments revised the Cyber Security Plan Implementation Milestone No. 8 completion date and the physical protection license condition.

Date of issuance: September 29, 2014.

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

Amendment Nos.: Unit 1-333 and Unit 2-326. A publicly-available version is in ADAMS under Accession No. ML14245A179; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-77 and DPR-79. The amendments revised the Operating License.

Date of initial notice in Federal Register: July 8, 2014 (79 FR 38582). The supplemental letter dated May 27, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2014.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 29, 2014. A redacted version was provided by letter dated May 27, 2014.

Brief description of amendment: The amendments revised the Cyber Security Plan Implementation Milestone No. 8 completion date and the physical protection license condition.Start Printed Page 64232

Date of issuance: September 29, 2014.

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

Amendment Nos.: Unit 1-333 and Unit 2-326. A publicly-available version is in ADAMS under Accession No. ML14245A179; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-77 and DPR-79. The amendments revised the Operating License.

Date of initial notice in Federal Register: July 8, 2014 (79 FR 38582). The supplemental letter dated May 27, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2014.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 29, 2014, as supplemented by letter dated May 27, 2014.

Brief description of amendment: The amendment revised the Cyber Security Plan Implementation Milestone No. 8 completion date and the physical protection license condition.

Date of issuance: September 29, 2014.

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

Amendment No.: 97. A publicly-available version is in ADAMS under Accession No. ML14255A152; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment.

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

Date of initial notice in Federal Register: July 8, 2014 (79 FR 38581). The supplemental letter dated May 27, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in the SE dated September 29, 2014.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 16th day of October 2014.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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

End Signature End Supplemental Information

[FR Doc. 2014-25357 Filed 10-27-14; 8:45 am]

BILLING CODE 7590-01-P