Skip to Content

Notice

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

Document Details

Information about this document as published in the Federal Register.

Enhanced Content

Relevant information about this document from Regulations.gov provides additional context. This information is not part of the official Federal Register document.

Published Document

This document has been published in the Federal Register. Use the PDF linked in the document sidebar for the official electronic format.

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 September 4, 2014 to September 17, 2014. The last biweekly notice was published on September 16, 2014.

DATES:

Comments must be filed by October 30, 2014. A request for a hearing must be filed by December 1, 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-2014-0207. 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.

Start Further Info

FOR FURTHER INFORMATION CONTACT:

Mable Henderson, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-3760, email: Mable.Henderson@nrc.gov.

End Further Info End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2014-0207 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-2014-0207.
  • 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-2014-0207 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.Start Printed Page 58813

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 which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

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 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://Start Printed Page 58814www.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 the NRC's guidance available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

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

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

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, 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.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant (PNP), Van Buren County, Michigan

Date of amendment request: July 29, 2014. A publicly-available version is in ADAMS under Accession No. ML14211A520.

Description of amendment request: The amendment would change the Operating License at PNP. Specifically, the amendment requests authorization to implement 10 CFR 50.61a, “Alternate fracture toughness requirements for protection against pressurized thermal shock events,” in lieu of 10 CFR 50.61, “Fracture toughness requirements for protection against pressurized thermal shock events.” PNP currently complies with 10 CFR 50.61. The 10 CFR 50.61 screening criteria define a limiting level of embrittlement beyond which plant operation cannot continue without further evaluation. As described in NUREG-1806, “Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61),” August 2007, the screening criteria in the PTS rule is overly conservative and the risk of through-wall cracking due to a PTS event is much lower than previously estimated. A publically-available version of NUREG-1806 is in ADAMS under Accession No. ML072830074.

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 Start Printed Page 58815consideration, 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.

This amendment request would allow implementation of the 10 CFR 50.61a alternate pressurized thermal shock (PTS) rule in lieu of the 10 CFR 50.61 PTS rule, and would not involve a significant increase in the probability or consequences of an accident. Application of 10 CFR 50.61a in lieu of 10 CFR 50.61 would not result in physical alteration of a plant structure, system or component, or installation of new or different types of equipment. Further, application of 10 CFR 50.61a would not significantly affect the probability of accidents previously evaluated in the Updated Final Safety Analysis Report (UFSAR) or cause a change to any of the dose analyses associated with the UFSAR accidents because accident mitigation functions would remain unchanged. Use of 10 CFR 50.61a would change how fracture toughness of the reactor vessel is assessed and does not affect reactor vessel neutron radiation fluence. As such, implementation of 10 CFR 50.61a in lieu of 10 CFR 50.61 would not increase the likelihood of a malfunction.

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 type of accident from any accident previously evaluated?

Response: No.

The amendment request would allow implementation of the 10 CFR 50.61a alternate PTS rule in lieu of 10 CFR 50.61. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. No physical plant alterations are made as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The amendment request would authorize implementation of 10 CFR 50.61a in lieu of 10 CFR 50.61. Regulation 10 CFR 50.61a would maintain the same functional requirements for the facility as 10 CFR 50.61. It establishes screening criteria that limit levels of embrittlement beyond which operation cannot continue without further plant-specific evaluation or modifications. Sufficient safety margins are maintained to ensure that any potential increases in core damage frequency and large early release frequency resulting from implementation of 10 CFR 50.61a are negligible. As such, there would be no significant reduction in the margin of safety as a result of use of the alternate PTS rule. The margin of safety associated with the acceptance criteria of accidents previously evaluated in the UFSAR is unchanged. The proposed change would have no effect on the availability, operability, or performance of the safety-related systems and components.

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: Mr. William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: David L. Pelton.

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. 50-010, 50-237 and 50-249, Dresden Nuclear Power Station, Units 1, 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: August 11, 2014. A publicly-available version is in ADAMS under Accession No. ML14224A245.

Description of amendment request: The proposed changes would revise the description for the Emergency Response Organization (ERO) requalification training frequency for Exelon personnel defined in Exelon's governing Emergency Plans for the named stations from annually to “once per calendar year not to exceed 18 months between training sessions.”

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:

Exelon has evaluated the proposed changes to the affected sites' Emergency Plans and determined that the changes do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards, set forth in 10 CFR 50.92, “Issuance of amendment,” is provided below.

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

Response: No.

The proposed changes do not increase the probability or consequences of an accident. The proposed changes do not involve the modification of any plant equipment or affect plant operation. The proposed changes will have no impact on any safety-related Structures, Systems, or Components (SSC).

The proposed changes would revise the ERO requalification frequency from an annual basis to once per calendar year not to exceed 18 months between training sessions defined in the Emergency Plan for the applicable Exelon facility. The proposed changes will align the Exelon legacy plants under one standard regarding the annual requalification training frequency for ERO personnel.

Therefore, the proposed changes to the Emergency Plan requalification training frequency for the affected sites do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes have no impact on the design, function, or operation of any plant SSC. The proposed changes do not affect plant equipment or accident analyses. The proposed changes only affect the administrative aspects of the annual ERO requalification training frequency requirements.

Therefore, the proposed changes to the Emergency Plan requalification training frequency for the affected sites do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There is no change being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. Margins of safety are unaffected by the proposed changes to the frequency in the ERO requalification training requirements.

Therefore, the proposed changes to the Emergency Plan requalification training frequency for the affected sites 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 Start Printed Page 58816standards 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: Bradley Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Travis L. Tate.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of application for amendments: July 25, 2014. A publicly-available version is in ADAMS under Accession No. ML14211A017.

Description of amendment request: The proposed amendment would change the definition in the PBAPS, Units 2 and 3, Technical Specifications (TS) for RECENTLY IRRADIATED FUEL. Specifically, the amendment would revise requirements pertaining to secondary containment hatches in order to facilitate activities performed during refueling outages.

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, with NRC staff revisions provided in [brackets], which is presented below:

1. Will operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to revise the PBAPS, Units 2 and 3, TS definition for RECENTLY IRRADIATED FUEL do not introduce new equipment or new equipment operating modes, nor do the proposed changes alter existing system relationships. The proposed changes do not affect plant operation, [any] design function, or any analysis that verifies the capability of a Structure, System, or Component (SSC) to perform a design function. There are no changes or modifications to [any] plant SSC. The plant Engineered Safety Features (ESFs) will continue to function as designed in all modes of operation. There are no significant changes to procedures or training being introduced by the proposed changes to the TS definition.

Based upon the results of the [fuel handling accident (FHA)] analysis, it has been demonstrated that, with the requested changes, the dose consequences remain within the regulatory guidance provided by the NRC as specified in 10 CFR 50.67 and associated Regulatory Guide (RG) 1.183 [ADAMS Accession No. ML003716792]. The calculations used to evaluate the consequences of the FHA accident in support of the proposed changes do not by themselves affect the plant response, but better represent the physical characteristics of the release, so that appropriate mitigation techniques may be applied. Therefore, the consequences of an accident previously evaluated are not significantly increased.

There is no adverse impact on systems designed to mitigate the consequences of accidents. The proposed changes do not adversely affect system or component pressures, temperatures, or flowrates for systems designed to prevent accidents or mitigate the consequences of an accident. Since these conditions are not adversely affected, the likelihood of failure of [an] SSC is not increased.

The proposed changes do not increase the likelihood of the malfunction of any SSC or impact any analyzed accident. Consequently, the probability or consequences of an accident previously evaluated are not affected.

Based on the above, Exelon concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes to revise the PBAPS, Units 2 and 3, TS definition for RECENTLY IRRADIATED FUEL do not alter the design function or operation of any SSC. There are no changes or modifications to [any] plant SSC. The plant ESFs will continue to function as designed. There is no new system component being installed, no new construction, and no performance of a new test or maintenance function. The proposed TS changes do not create the possibility of a new credible failure mechanism or malfunction. The proposed changes do not introduce new accident initiators or precursors of a new or different kind of accident. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed changes. [Secondary containment] integrity is not adversely impacted and radiological consequences from the analyzed FHA remain within specified regulatory limits. The proposed changes do not adversely impact system or component pressures, temperatures, or flowrates for systems designed to prevent accidents or mitigate the consequences of an accident. Since these conditions are not adversely impacted, the likelihood of failure of [an] SSC is not increased. Consequently, the proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

Based on the above, Exelon concludes that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes to revise the PBAPS, Units 2 and 3, TS definition for RECENTLY IRRADIATED FUEL do not alter the design function or operation of any SSC. There are no changes or modifications to [any] plant SSC. The plant ESFs will continue to function as designed. The proposed changes do not increase system or component pressures, temperatures, or flowrates for systems designed to prevent accidents or mitigate the consequences of an accident.

Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed event has been evaluated and margin has been retained to ensure that the analysis adequately bounds the postulated FHA event. The dose consequences resulting from analyzing the FHA design basis accident comply with the requirements of 10 CFR 50.67 and the guidance of RG 1.183.

The proposed changes continue to ensure that the doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) boundary, as well as the Main Control Room (MCR), remain within corresponding regulatory limits.

Based on the above, Exelon concludes 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: J. Bradley Fewell, Esquire, Vice President and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348.

Acting NRC Branch Chief: Robert G. Schaaf.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit 2, (BVPS-2) Beaver County, Pennsylvania

Date of amendment request: June 2, 2014, as supplemented by letter dated August 8, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14153A388, and ML14223A540, respectively.

Description of amendment request: The amendment would change the BVPS-2 technical specifications (TSs). Specifically, the proposed license amendment would revise TS 4.3.2, “Drainage,” to correct the minimum drain elevation for the spent fuel storage pool specified in the TS. In accordance with 10 CFR Part 50, Appendix B, Section XVI, “Corrective Action,” the proposed amendment is required to resolve a TS discrepancy regarding an existing plant design feature.

Basis for proposed no significant hazards consideration determination: Start Printed Page 58817As 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.

Previously evaluated accidents including a fuel handling accident and spent fuel cask drop accident are not affected by the proposed amendment. Reducing the minimum water level above fuel stored in the spent fuel storage pool in the event of inadvertent draining as proposed would not involve a significant increase in the probability of a previously evaluated accident. Maloperation or passive piping failure causing inadvertent draining of the spent fuel storage pool is not postulated concurrent with the fuel handling or spent fuel cask drop accident. The proposed amendment would not result in any failure modes that could initiate an analyzed accident, and does not increase the likelihood of a malfunction of a system, structure or component; therefore, the probability of analyzed accidents is not affected.

There are no changes to how the station will be operated, limiting conditions for operation, or limiting safety system settings. The proposed amendment does not affect the capability of a system, structure or component to perform a design function. Since design functions are not affected by the proposed amendment, the consequences of previously evaluated accidents 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.

Reducing the minimum water level above fuel stored in the spent fuel storage pool in the event of inadvertent draining as proposed does not create any new failure mechanisms, malfunctions, or accident initiators and does not change design functions or system operation in a way that affects the ability of systems, structures, and components to perform design functions.

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.

General Design Criterion 61, “Fuel storage and handling and radioactivity control,” of 10 CFR 50, Appendix A, states in part that fuel storage and handling systems shall be designed with suitable shielding for radiation protection.

The proposed change involves a reduction in the minimum elevation of piping and penetrations of the spent fuel storage pool specified in the Technical Specifications. In the event maloperation or passive piping failure causes inadvertent draining of the spent fuel storage pool, the remaining water level in the pool ensures the stored fuel remains covered, provides adequate shielding for personnel, and affords adequate assurance of safety when judged against the current regulatory standard of General Design Criterion 61.

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

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

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

Acting NRC Branch Chief: Robert G. Schaaf.

FirstEnergy Nuclear Operating Company (FENOC), Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Perry, OH

Date of amendment request: June 23, 2014. A publicly-available version is in ADAMS under Accession No. ML14174A633.

Description of amendment request: The proposed amendment updates the technical specification (TS) pressure and temperature (P/T) figures using an NRC approved methodology to adjust the P/T limit curves for previously missing data, addresses the reactor coolant system (RCS) vacuum condition that can occur under certain conditions, and aligns the heatup/cooldown requirements of the TS with the limits in the associated P/T figures. Additionally editorial changes are proposed related to the P/T figures including clarifications and updates to the associated titles, labeling, and notes.

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 any accident previously evaluated?

Response: No.

The P/T [pressure and temperature] limits define RCS [reactor coolant system] operational limits to avoid encountering pressure, temperature, and temperature rate of change conditions that reduce safety margins with respect to nonductile brittle failure of the reactor coolant pressure boundary (RCPB). The figures are not accident initiators or accident mitigating features, but preclude operation in an unanalyzed condition.

This proposed amendment does not change the design function of the RCS or RCPB and does not change the way the plant is maintained or operated when using the P/T limit curves. This proposed amendment does not affect any plant systems that are accident initiators and does not affect any accident mitigating feature.

The proposed amendment does not affect the operability requirements for the RCS, as verification of operating within the P/T limits will continue to be performed, as required. Compliance with and continued verification of the P/T limits support the capability of the RCS to perform its required design functions, consistent with the plant safety analyses.

Changing the figures will not change any of the dose analyses associated with the USAR [updated safety analysis report] Chapter 15 accidents because they do not affect the source term, containment isolation or radiological release assumptions used in any accident previously evaluated. Plant accident mitigation functions and requirements remain unchanged.

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 P/T limits define RCS operational parameters to protect the RCPB and are not accident initiators or accident mitigating features. The limits are conservatively calculated using an NRC approved methodology. This proposed amendment does not change the design function of the RCS or RCPB, and does not change the way the plant is operated or maintained. This proposed amendment does not affect any plant systems that are accident initiators, does not affect any accident mitigating feature, and does not create 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 change involve a significant reduction in the margin of safety?

Response: No.

The P/T limits define RCS operational parameters, which are established to protect the reactor vessel. The analysis supporting the curve changes utilize methods previously reviewed and approved by the NRC.

Margin of safety is related to the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. This proposed amendment does not directly involve or physically affect fuel cladding or the primary containment.

The amendment request proposes to update the P/T limit figures using an NRC approved methodology. The curves maintain the margin of safety for RCPB materials that are exposed to neutron radiation.Start Printed Page 58818

The proposed amendment does not involve a physical change to the plant, does not change methods of plant operation within prescribed limits, and does not change methods of maintenance on equipment important to safety. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the responses to the three questions above, FENOC [FirstEnergy Nuclear Operating Company] concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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, Attorney, FirstEnergy Corporation, Mail Stop. A-GO-15, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Travis L. Tate.

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

Date of amendment request: June 9, 2014. A publicly-available version is in ADAMS under Accession No. ML14175A121.

Description of amendment request: The amendments would revise Technical Specification (TS) 6.2, Organization, specifically TS 6.2.2.e. to allow the station technical assistant (STA) position to be manned by a single STA, a shift supervisor who meets the qualifications for the STA, or an individual with a senior reactor operator's license who meets the qualifications for the STA on each unit in MODES 1, 2, 3, or 4. This criterion was omitted from FPL's license amendment request dated July 26, 2013 (ADAMS Accession No. ML13219A840), that addressed shift staffing requirements. As a result, it was omitted from the corresponding license amendments dated February 7, 2014 (ADAMS Accession No. ML14016A248). This criterion was previously approved by the NRC and incorporated into the St. Lucie Units 1 and 2 TSs by Amendment Nos. 173 and 113, respectively.

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 changes define the method for manning the shift technical advisor (STA) position and do not reduce the unit staffing requirements. In addition, the changes correct a typographical error. The changes do not affect the minimum shift compliment in any mode of operation nor decrease the effectiveness of shift personnel. The STA position will continue to be manned by qualified personnel. The proposed changes are administrative and editorial in nature and will not result in any significant increase in the probability of consequences of an accident as previously evaluated. Further, the proposed changes do not introduce additional risk or greater potential for consequences of an accident that has not previously been evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes define the method for manning the shift technical advisor position and do not reduce the unit staffing requirements. In addition, the changes correct a typographical error. The proposed changes are administrative and editorial in nature. No new or different type of equipment will be installed. The proposed changes will not introduce new failure modes/effects that could lead to an accident for which consequences exceed that of accidents previously analyzed. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes define the method for manning the STA position and do not reduce the unit staffing requirements. In addition, the changes correct a typographical error. The changes do not affect the minimum shift compliment in any mode of operation nor decrease the effectiveness of shift personnel. The STA position will continue to be manned by qualified personnel. The proposed changes will not involve a significant reduction in a margin of safety in that the changes are administrative and editorial in nature. No plant equipment or accident analyses will be affected. Additionally, the proposed changes will not relax any criteria used to establish safety limits, safety system settings, or the bases for any limiting conditions for operation. Safety analysis acceptance criteria are not affected. Plant operation will continue within the design basis.

The proposed changes do not adversely affect systems that respond to safely shutdown the plant, and maintain the plant in a safe shutdown condition. Consequently, 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 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.

Acting NRC Branch Chief: Lisa M. Regner.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, Unit 2, St. Lucie County, Florida

Date of amendment request: January 30, 2014. A publicly-available version is in ADAMS under Accession No. ML14049A284.

Description of amendment request: The amendment would revise the Technical Specification (TS) surveillance requirements (SRs) for snubbers to conform to revisions to the Snubber Testing Program allowing a year extension to the existing interval for the snubber program transition. This revision would meet the requirements of the Operation and Maintenance (OM) Code and Subsection ISTD, “Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light Water Reactor Nuclear Power Plants,” of the American Society of Mechanical Engineers OM Code, 2004 Edition with 2005 and 2006 Addenda.

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 changes would revise SR 4.7.9 to conform the TS to the revised surveillance program for snubbers. Snubber examination, testing and service life monitoring will continue to meet the requirements of 10 CFR 50.55a(g). Snubber examination, testing and service life monitoring is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. Snubbers will continue to be demonstrated OPERABLE by performance of a Start Printed Page 58819program for examination, testing and service life monitoring in compliance with 10 CFR 50.55a or authorized alternatives. The proposed change to TS ACTION 3.7.9 for inoperable snubbers is administrative in nature and is required for consistency with the proposed change to SR 4.7.9. The proposed change does not adversely affect plant operations, design functions or analyses that verify the capability of systems, structures, and components to perform their design functions therefore, the consequences of accidents previously evaluated are not significantly increased. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes do not involve any physical alteration of plant equipment. The proposed changes do not alter the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

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

The proposed change to TS ACTION 3.7.9 for inoperable snubbers is administrative in nature and is required for consistency with the proposed change to SR 4.7.9. Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's 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.

Acting NRC Branch Chief: Lisa M. Regner.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station (CNS), Nemaha County, Nebraska

Date of amendment request: July 14, 2014. A publicly-available version is in ADAMS under Accession No. ML14202A205.

Description of amendment request: The proposed amendment would delete Technical Specification 5.5.3, “Post Accident Sampling,” thereby eliminating the program requirements to have and maintain the post-accident sampling system. The changes are consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-413, “Elimination of Requirements for a Post Accident Sampling System (PASS).” The availability of this technical specification improvement was announced in the Federal Register on March 20, 2002, as part of the consolidated line item improvement process. CNS will continue to have the ability to obtain samples, utilizing PASS, following an accident.

Basis for proposed no significant hazards consideration determination: The licensee stated in its application that it reviewed the proposed no significant hazards consideration determination published on December 27, 2001 (66 FR 66949), as part of the consolidated line item improvement process. The licensee stated that it concluded that the proposed determination presented in the notice is applicable to CNS and the determination is incorporated by reference to satisfy the requirements of 10 CFR 50.91(a). As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

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

The PASS was originally designed to perform many sampling and analysis functions. These functions were designed and intended to be used in post accident situations and were put into place as a result of the TMI-2 [Three Mile Island, Unit 2] accident. The specific intent of the PASS was to provide a system that has the capability to obtain and analyze samples of plant fluids containing potentially high levels of radioactivity, without exceeding plant personnel radiation exposure limits. Analytical results of these samples would be used largely for verification purposes in aiding the plant staff in assessing the extent of core damage and subsequent offsite radiological dose projections. The system was not intended to and does not serve a function for preventing accidents and its elimination would not affect the probability of accidents previously evaluated.

In the 20 years since the TMI-2 accident and the consequential promulgation of post accident sampling requirements, operating experience has demonstrated that a PASS provides little actual benefit to post accident mitigation. Past experience has indicated that there exists in-plant instrumentation and methodologies available in lieu of a PASS for collecting and assimilating information needed to assess core damage following an accident. Furthermore, the implementation of Severe Accident Management Guidance (SAMG) emphasizes accident management strategies based on in-plant instruments. These strategies provide guidance to the plant staff for mitigation and recovery from a severe accident. Based on current severe accident management strategies and guidelines, it is determined that the PASS provides little benefit to the plant staff in coping with an accident.

The regulatory requirements for the PASS can be eliminated without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. The elimination of the PASS will not prevent an accident management strategy that meets the initial intent of the post-TMI-2 accident guidance through the use of the SAMGs, the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

Therefore, the elimination of PASS requirements from Technical Specifications (TS) (and other elements of the licensing bases) does not involve a significant increase in the consequences of any accident previously evaluated.

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

The elimination of PASS related requirements will not result in any failure mode not previously analyzed. The PASS was intended to allow for verification of the extent of reactor core damage and also to provide an input to offsite dose projection calculations. The PASS is not considered an accident precursor, nor does its existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building.Start Printed Page 58820

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

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

The elimination of the PASS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. Methodologies that are not reliant on PASS are designed to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The use of a PASS is redundant and does not provide quick recognition of core events or rapid response to events in progress. The intent of the requirements established as a result of the TMI-2 accident can be adequately met without reliance on a PASS.

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

The NRC staff has reviewed the 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: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

Acting NRC Branch Chief: Eric R. Oesterle.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: July 17, 2014. A publicly-available version is in ADAMS under Accession No. ML14203A045.

Description of amendment request: The proposed amendment would move the Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from the Technical Requirements Manual (TRM) to the Technical Specifications (TS).

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. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

LHGR limits have been defined to provide sufficient margin between the steady-state operating condition and any fuel damage condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time. The proposed change to move the LHGR limits from the TRM to TS, including the change to TS 3.4.1, Recirculation Loops Operating, and TS 3.7.7, Main Turbine Bypass System, does not modify the limits, change assumptions for the accident analysis, or change operation of the station.

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 modify the limits, change assumptions for the accident analysis, or change operation of the station.

The proposed change does move LHGR limits that have been defined to provide sufficient margin between the steady-state operating condition and any fuel damage condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time from the TRM to TS.

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 to move the LHGR limits from the TRM to TS, including the change to TS 3.4.1 and TS 3.7.7, does not modify the limits, change assumptions for the accident analysis, or change operation of the station.

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: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

Acting NRC Branch Chief: Eric R. Oesterle.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of amendment request: June 23, 2014. A publicly-available version is in ADAMS under Accession No. ML14175B387.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) requirements to address NRC Generic Letter (GL) 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” as described in Technical Specification Task Force (TSTF) Change Traveler TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”

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 [Surveillance Requirements] that require verification that the Emergency Core Cooling Systems (ECCS), Residual Heat Removal (RHR) System, and the Reactor Core Isolation Cooling (RCIC) 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 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.

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, and RCIC 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, and RCIC System are not rendered inoperable due to accumulated gas and to Start Printed Page 58821provide 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.

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: Mr. James Petro, P. O. Box 14000 Juno Beach, FL 33408-0420.

NRC Branch Chief: David L. Pelton.

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

Date of amendment request: July 24, 2014. A publicly-available version is in ADAMS under Accession No. ML14209A918.

Description of amendment request: The proposed amendment would revise Seabrook Technical Specifications (TSs) by increasing the voltage limit for a full load rejection test of the emergency diesel generator specified in surveillance requirement 4.8.1.1.2.f.3 of TS 3.8.1.1, “A.C. [alternating current] Sources—Operating.” The proposed amendment also revises the TS definition of the terms “Operable—Operability.”

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, with NRC staff revisions provided in [brackets], 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 to increase in the [emergency diesel generator] EDG full load rejection overvoltage limit from 4784 [volts] V to 4992V is not an accident initiator. The overvoltage transient is an expected response to a full load rejection. The magnitude and duration of the proposed overvoltage limit have been considered and determined to have no detrimental effects on the connected equipment that is exposed to the voltage transient. The proposed change does not affect the EDG design function or how the EDG is operated. Since the EDG is not impacted, the EDG remains capable of performing its intended design function of supplying power to emergency safeguards equipment. The proposed change to the definition of operable—operability is administrative in nature and does not alter the meaning of the defined terms.

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

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

Response: No.

The proposed changes to revise the definition of the terms operable—operability and to increase the EDG full load rejection overvoltage limit from 4784V to 4992V are not accident initiators. The overvoltage transient is an expected response to a full load rejection. The magnitude and duration of the proposed overvoltage limit have been considered and determined to have no detrimental effects on the connected equipment that is exposed to the voltage transient. The proposed changes do not introduce any new failure modes.

The changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change in the methods for operating the plant. The proposed changes do not affect the EDG design function or how the EDG is operated. Since the EDG is not impacted, the EDG remains capable of performing its intended design function of supplying power to emergency safeguards equipment. The change to the definition of operable—operability makes grammatical corrections and adds clarity but makes no change to the meaning of the terms.

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

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

Response: No.

The proposed change to increase in the EDG full load rejection overvoltage limit from 4784V to 4992V has been evaluated with consideration of the effect on the EDG and connected equipment that would be exposed to the higher voltage transient. Based on review of equipment specifications, test data, and manufacturer's input, it was concluded that there would be no detrimental effects to the EDG or connected equipment that is exposed to the higher voltage transient. The EDG remains capable of performing its intended design function of supplying power to emergency safeguards equipment.

The proposed change to the definition of operable—operability is administrative in nature and does not alter any criterion used to establish operability of plant structure, systems, or components.

The proposed amendment does not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings. The changes do not adversely impact plant operating margins or the reliability of equipment credited in the safety analyses.

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

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

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

Acting NRC Branch Chief: Robert G. Schaaf.

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

Date of amendment request: July 24, 2014. A publicly-available version is in ADAMS under Accession No. ML14209A919.

Description of amendment request: The proposed amendment would revise the Seabrook Technical Specifications (TS). The proposed change modifies TS 3.3.3.1, “Radiation Monitoring for Plant Operations,” to eliminate duplicate requirements, resolve an inconsistency, and correct a deficiency.

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, with NRC staff revisions provided in [brackets], 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 instruments involved with the proposed changes to the technical specifications (TS) are not initiators of any accidents previously evaluated, and the probability and consequences of accidents previously evaluated are unaffected by the proposed changes. There is no change to any equipment response or accident scenario, and the changes impose no additional challenges to fission product barrier integrity. The proposed changes do not alter the design, function, operation, or configuration of any plant structure, system, or component (SSC). As a result, the outcomes of accidents previously evaluated are unaffected. The Start Printed Page 58822proposed changes modify the TS to eliminate duplicate requirements, resolve an inconsistency, and correct a deficiency.

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

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

Response: No.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The changes do not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, physical configuration, or method of operation of any plant SSC. No physical changes are made to the plant, so no new causal mechanisms are introduced.

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

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

Response: No.

The ability of any operable SSC to perform its designated safety function is unaffected by the proposed changes. The proposed changes do not alter any safety analyses assumptions, safety limits, limiting safety system settings, or method of operating the plant. The changes do not adversely impact plant operating margins or the reliability of equipment credited in the safety analyses.

Therefore, the proposed change 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: William Blair, Managing Attorney, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

Acting NRC Branch Chief: Robert G. Schaaf.

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

Date of amendment request: July 24, 2014. A publicly-available version is in ADAMS under Accession No. ML14216A404.

Description of amendment request: The proposed amendment would incorporate revised reactor coolant system (RCS) pressure-temperature limits in the Technical Specification (TS) applicable to 55 effective full-power years. The change will also provide new overpressure protection setpoints and lower the RCS temperature at which the TS is applicable.

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 changes to the Technical Specifications (TS) do not impact the physical function of plant structures, systems, or components (SSCs) or the manner in which SSCs perform their design function. Operation in accordance with the proposed TS will ensure that all analyzed accidents will continue to be mitigated by the SSCs as previously analyzed. The proposed changes do not alter or prevent the ability of operable SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions.

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

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

Response: No.

The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), create new failure modes for existing equipment, or create any new limiting single failures. The changes to the pressure—temperature limits, power operated relief valve setpoints, and the over pressure protection system effective temperature will continue to ensure that appropriate fracture toughness margins are maintained to protect against reactor vessel failure, during both normal and low temperature operation. The proposed changes are consistent with the applicable NRC approved methodologies (i.e., WCAP-14040, Rev. 4 and ASME Code Case N-641). Plant operation will not be altered, and all safety functions will continue to perform as previously assumed in accident analyses.

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

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

Response: No.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes will not adversely affect the operation of plant equipment or the function of any equipment assumed in the accident analysis. The proposed changes were developed using NRC approved methodologies and will continue to ensure an acceptable margin of safety is maintained. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

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

Acting NRC Branch Chief: Robert G. Schaaf.

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

Date of amendment request: June 17, 2014. A publicly-available version is in ADAMS under Accession No. ML14168A486.

Description of amendment request: The NSPM proposes to revise MNGP Technical Specification (TS) 3.5.1, “ECCS [Emergency Core Cooling System]—Operating,” to correct the requirements for the Alternate Nitrogen System pressure. TS Surveillance Requirement (SR) 3.5.1.3 requires verification of limits for automatic depressurization system (ADS) pneumatic pressure for both ADS pneumatic supplies. The proposed change would revise the TS SR 3.5.1.3.b pressure limit for determining operability of the Alternate Nitrogen System from greater than or equal to (≥) 410 pounds per square inch gauge (psig) to a corrected value of ≥ 700 psig.

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 provided 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 the TS SR for the purpose of restoring a value to be consistent with the licensing basis. The proposed TS change does not introduce new Start Printed Page 58823equipment or new equipment operating modes, nor does the proposed change alter existing system relationships. The proposed change does not affect plant operation, design function or any analysis that verifies the capability of a system, structure or component (SSC) to perform a design function. Further, the proposed change does not increase the likelihood of the malfunction of any SSC or impact any analyzed accident. Consequently, the probability of an accident previously evaluated is not affected and there is not significant increase in the consequences of any accident previously evaluated.

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.

There proposed change revises the TS SR for the purpose of restoring a value to be consistent with the licensing basis. The change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operations. The proposed change does not alter assumptions made in the safety analysis for the components supplied by the Alternate Nitrogen System. Further, the proposed change does not introduce new accident initiators.

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 the TS SR for the purpose of restoring a value to be consistent with the licensing basis. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis assumptions and acceptance criteria are not affected by this change.

Therefore, the proposed change 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: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David L. Pelton.

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

Date of amendment request: July 28, 2014. A publicly-available version is in ADAMS under Accession No. ML14209B074.

Description of amendment request: The proposed amendments would modify the technical specifications (TS) to risk-inform requirements regarding selected Required Action End States. The proposed changes to the Required Action End States are described in Table 1 of the Enclosure to the licensee's letter dated July 28, 2014. The changes are consistent with Technical Specification Task Force (TSTF) Traveler TSTF-432, Revision 1, “Change in Technical Specifications End States (WCAP-16294)” (ADAMS Accession No. ML103430249).

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 modifies the end state (e.g., mode or other specified condition) which the Required Actions specify must be entered if compliance with the Limiting Conditions for Operation (LCO) is not restored. The requested Technical Specifications (TS) permit an end state of Mode 4 rather than an end state of Mode 5 contained in the current TS. In some cases, other Conditions and Required Actions are revised to implement the proposed change. Required Actions are not an initiator of any accident previously evaluated. Therefore, the proposed change does not affect the probability of any accident previously evaluated. The affected systems continued to be required to be operable by the TS and the Completion Times specified in the TS to restore equipment to operable status or take other remedial Actions remain unchanged. WCAP-16294-NP-A, Revision 1, “Risk-Informed Evaluation of Changes to [Technical Specification] Required Action Endstates for Westinghouse NSSS [Nuclear Steam Supply System] PWRs [Pressurized Water Reactors],” demonstrates that the proposed change does not significantly increase the consequences of any accident previously evaluated.

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

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

Response: No.

The proposed change modifies the end state (e.g., mode or other specified condition) which the Required Actions specify must be entered if compliance with the LCO is not restored. In some cases, other Conditions and Required Actions are revised to implement the proposed change. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the change does not impose any new requirements. The change does not alter assumptions made in the safety analysis.

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

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

Response: No.

The proposed change modifies the end state (e.g., mode or other specified condition) which the Required Actions specify must be entered if compliance with the LCO is not restored. In some cases, other Conditions and Required Actions are revised to implement the proposed change. Remaining within the Applicability of the LCO is acceptable because WCAP-16294-NP-A demonstrates that the plant risk in MODE 4 is similar to, or lower than, MODE 5. As a result, no margin of safety is significantly affected.

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 requests involve no significant hazards consideration.

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

Acting NRC Branch Chief: Eric R. Oesterle.

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 12, 2014. A publicly-available version is in ADAMS under Accession No. ML14164A098.

Description of amendment request: The proposed license amendment request (LAR) proposes to revise Plant Specific Tier 2* material within the Updated Final Safety Analysis Report (UFSAR) by making editorial and consistency corrections.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the Start Printed Page 58824issue 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 editorial and consistency update does not involve a technical change, i.e., there is no design parameter or requirement, calculation, analysis, function, or qualification change. No structure, system, component (SSC), design, or function would be adversely affected. No design or safety analysis would be adversely affected. The proposed changes do not adversely affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not adversely affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not adversely affected.

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

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

Response: No.

The proposed editorial and consistency update would not affect the design or function of any SSC, but will instead provide consistency between the SSC designs and functions and the discussions currently presented in the UFSAR via Tier 2* information. The proposed nontechnical changes would not introduce a new failure mode, 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 from any accident previously evaluated.

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

Response: No.

The proposed editorial and consistency update is nontechnical and thus would not affect any design parameter, function, or analysis. There would be no change to an existing design basis, design function, regulatory criterion, or analyses. No safety analysis or design basis acceptance limit/criterion is involved.

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

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

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

NRC Branch Chief: Lawrence J. Burkhart.

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: August 28, 2014. A publicly-available version is in ADAMS under Accession No. ML14245A601.

Description of amendment request: The proposed license amendment request (LAR) proposes to revise Tier 2* and Tier 2 information related to the design details of connections in several locations between the steel plate composite construction (SC) used for the shield building and the standard reinforced concrete (RC) walls, floors, and roofs of the auxiliary building and the lowers of the shield building.

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

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

Response: No.

The design function of the nuclear island structures is to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in the nuclear island. The nuclear island structures are structurally designed to meet seismic Category I requirements as defined in Regulatory Guide 1.29.

The changes to the detail design of connections between the RC and SC structures do not have an adverse impact on the response of the nuclear island structures to safe shutdown earthquake ground motions or loads due to anticipated transients or postulated accident conditions. The changes to the detail design do not impact the support, design, or operation of mechanical and fluid systems. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions. The plant response to previously evaluated accidents or external events is not adversely affected, nor do the changes described create any new accident precursors.

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

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

Response: No.

The proposed changes are to the detail design of connections between the RC and SC structures. The changes to the detail design of connections do not change the criteria and requirements for the design and analysis of the nuclear island structures. The changes to the detail design of connections do not change the design function, support, design, or operation of mechanical and fluid systems. The changes to the detail design of connections do not change the methods used to connect the RC to SC. The changes of the detail design of connections do not result in a new failure mechanism for the nuclear island structures or new accident precursors. As a result, the design functions of the nuclear island structures are not adversely affected by the proposed changes.

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.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes; and thus, no margin of safety is reduced.

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

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

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

NRC Branch Chief: Lawrence J. Burkhart.

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

Date of amendment request: July 29, 2014. A publicly-available version is available in ADAMS under Accession No. ML14210A646.

Description of amendment request: The proposed license amendment request would revise the Combined Licenses (COLs) with regard to Tier 1 material and promote consistency with the Updated Final Safety Analysis Report Tier 2.

Nuclear Operating Company has also requested an exemption from the provisions of 10 CFR part 52, appendix D, section III.B, “Design Certification Rule for the AP1000 Design, Scope and Contents,” to allow a departure from the elements of the certification information in Tier 1 of the generic Design Control Document.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the Start Printed Page 58825licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No

The proposed editorial and consistency COL Appendix C and corresponding plant-specific Tier 1 update does not involve a technical change, e.g., there is no design parameter or requirement, calculation, analysis, function or qualification change. No structure, system, or component (SSC) design or function would be affected. No design or safety analysis would be affected. The proposed changes do not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not affected.

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

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

Response: No

The proposed editorial and consistency COL Appendix C and corresponding plant-specific Tier 1 update would not affect the design or function of any SSC, but will instead provide consistency between the SSC designs and functions currently presented in the UFSAR, COL Appendix C, and the Tier 1 information. The proposed changes would not introduce a new failure mode, 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.

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

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

Response: No

The proposed editorial and consistency COL Appendix C and corresponding plant-specific Tier 1 update would not affect the design or function of any SSC, but will instead provide consistency between the SSC designs and functions currently presented in the UFSAR, COL Appendix C, and the Tier 1 information. The proposed changes would not introduce a new failure mode, fault or sequence of events that could result in a radioactive material release. Therefore, the requested amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Lawrence Burkhart.

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

Date of amendment request: July 30, 2014. A publicly-available version is in ADAMS under Accession No. ML14211A666.

Description of amendment request: The proposed license amendment request would revise the combined licenses (COLs) with regard to Tier 2* material within the Updated Final Safety Analysis Report (UFSAR) to resolve inconsistencies with other Tier 2* information elsewhere in the UFSAR.

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 requested amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed editorial and consistency update does not involve a technical change, i.e., there is no design parameter or requirement, calculation, analysis, function, or qualification change. No structure, system, or component, design, or function would be adversely affected. No design or safety analysis would be adversely affected. The proposed changes do not adversely affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not adversely affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved, thus the radiological releases in the accident analyses are not adversely affected.

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

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

Response: No.

The proposed editorial and consistency update would not affect the design or function of any structure, system, or component, but will instead provide consistency between the structure, system, and component designs and functions and the discussions currently presented in the UFSAR via Tier 2* information. The proposed non-technical changes would not introduce a new failure mode, fault, or sequence of events that could result in a radioactive material release.

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

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

Response: No.

The proposed editorial and consistency update is non-technical and thus would not affect any design parameter, function, or analysis. There would be no change to an existing design basis, design function, regulatory criterion, or analyses. No safety analysis or design basis acceptance limit/criterion is involved.

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

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

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

NRC Branch Chief: Lawrence Burkhart.

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

Date of amendment request: April 11, 2014. A publicly-available version is in ADAMS under Accession No. ML14101A459. The amendment request was supplemented by letter dated April 18, 2014. A publicly-available version is in ADAMS under Accession No. ML14108A093. The amendment request was further supplemented by two letters dated August 28, 2014. Publicly-available versions of the two letters are in ADAMS under Accession Nos. ML14241A250 and ML14241A264.

Description of amendment request: The license amendment request was originally noticed in the Federal Register on June 6, 2014 (79 FR 32771). This notice is being reissued in its entirety to include the revised analysis of the issue of no significant hazards consideration submitted by the licensee in its August 28, 2014, submission (ADAMS Accession No. ML14241A250). The proposed license amendment request would depart from the plant-specific Design Control Document Tier 1 and Tier 2 material to describe modifications to increase the efficiency of the return of condensate utilized by the passive core cooling system to the Start Printed Page 58826in-containment refueling water storage tank to support the capability for long term cooling.

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 containment condensate flow path changes provide sufficient condensate return flow to maintain In-containment Refueling Water Storage Tank (IRWST) level above the top of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) tubes long enough to prevent PRHR HX performance degradation from that considered in the UFSAR [Updated Final Safety Analysis Report] Chapter 15 safety analyses. The added components are seismically qualified and constructed of only those materials appropriately suited for exposure to the reactor coolant environment as described in UFSAR Section 6.1. No aluminum is permitted to be used in the construction of these components so that they do not contribute to hydrogen production in containment.

The proposed changes clarify the design basis for the PRHR HX, which removes decay heat from the Reactor Coolant System (RCS) during a non-loss of coolant accident (non-LOCA). With operator action to avoid unnecessary Automatic Depressurization System (ADS) actuation based on RCS conditions, PRHR HX operation can be extended longer than would be maintained automatically by the protection system. Though analysis shows significantly greater capacity, the extent of the capability of the PRHR HX would be changed from operating indefinitely to operating for at least 72 hours. If PRHR HX capability were exhausted after 72 hours, the ADS would be actuated, which could result in significant containment floodup. However, probabilistic analysis shows the probability of design basis containment floodup after PRHR HX operation during a non-LOCA event is significantly lower than the probability of a small break LOCA, for which comparable containment floodup is anticipated.

Therefore, the probability of significant containment floodup is not increased.

The proposed changes do not affect any components whose failure could initiate a previously evaluated event, thus the probabilities of the accidents previously evaluated are not affected. The affected equipment does not adversely affect or interact with safety-related equipment or another radioactive material barrier. The proposed changes clarify the post-accident performance requirements for the PRHR HX. However, the proposed changes do not prevent the engineered safety features from performing their safety-related accident mitigating functions. The radioactive material source terms and release paths used in the safety analyses are unchanged, thus the radiological releases in the UFSAR accident analyses are not affected.

Therefore, the proposed amendment does not involve an 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 long-term safe shutdown analysis results show that the PRHR HX continues to meet its acceptance criterion, i.e., to cool the Reactor Coolant System (RCS) to below 420ºF in 36 hours. The added equipment does not adversely interface with any component whose failure could initiate an accident, or any component that contains radioactive material. The modified components do not incorporate any active features relied upon to support normal operation. The downspout and gutter return components are seismically qualified to remain in place and functional during seismic and dynamic events. The containment condensate flow path changes do not create a new fault or sequence of events that could result in a radioactive material release.

The proposed change quantifies the duration that the PRHR HX is capable of maintaining adequate core cooling, and specifies that if PRHR HX cooling capability is exhausted, the ADS would be actuated. This involves the possibility of opening the ADS valves after the IRWST water level has decreased below the spargers, which promote steam condensation in the IRWST. During this condition, the loads on the IRWST, spargers and any internal structures or components in the IRWST would still be less than their limiting loads, and these SSCs would not be adversely affected or cause a different mode of operation. Therefore, no new type of accident could be created by this condition.

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 proposed changes do not reduce the redundancy or diversity of any safety-related function. The added components are classified as safety-related, seismically qualified, and are designed to comply with applicable design codes. The proposed containment condensate flow path changes provide sufficient condensate return flow to maintain adequate IRWST water level for those events using the PRHR HX cooling function. The long-term Shutdown Temperature Evaluation results in UFSAR Chapter 19E show the PRHR HX continues to meet its acceptance criterion. The UFSAR Chapters 6 and 15 analyses results are not affected, thus margins to their regulatory acceptance criteria are unchanged. The former design basis, which stated the PRHR HX could bring the plant to 420°F within 36 hours, is changed to state the heat exchanger can establish safe, stable conditions in the reactor coolant system after a design basis event. Such safe stable conditions may not coincide with an RCS temperature of 420°F. However, the PRHR HX is able to bring the RCS to a sufficiently low temperature such that RCS conditions would be comparable to those achieved at 420°F—peak cladding temperatures, departure from nucleate boiling, and pressurizer level would be maintained within acceptable limits of the evaluation criteria. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus no margin of safety is significantly reduced.

Therefore, the proposed amendment does not reduce 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: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

NRC Branch Chief: Lawrence J. Burkhart.

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

Date of amendment request: July 14, 2014. A publicly-available version is in ADAMS under Accession No. ML14195A296.

Description of amendment request: The proposed change would amend Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4 to modify the fire area fire barriers of the turbine building switchgear rooms on Elevations 141'-3” and 158'-7” of the turbine building to accommodate the revised layout of the low and medium voltage switchgear and associated equipment. The proposed changes also provide an editorial change to a fire area number. The requested amendment requires changes to Updated Final Safety Analysis Report (UFSAR) information, which include changes to plant-specific Tier 2* information and changes to Tier 2 information that involve changes to this 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.Start Printed Page 58827

The proposed reconfiguration of the turbine building switchgear rooms, the control system cabinet room, the new electrical equipment room, and the associated heating, ventilation and air conditioning (HVAC) room and the proposed editorial change would not adversely affect any safety-related equipment or function. The modified configuration will maintain the fire protection function (i.e., barrier) as evaluated in Updated Final Safety Analysis Report (UFSAR) Appendix 9A, thus, the probability of a spread of a fire from these areas 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 affect nonsafety-related electrical switchgear and do not involve any accident, initiating event, or component failure; thus, the probabilities of the accidents previously evaluated are not affected. The proposed changes do not interface with or affect any system containing radioactivity or affect any radiological material release source terms; 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 proposed changes to the fire zones in the turbine building related to the turbine building switchgear rooms, the control system cabinet room, the new electrical equipment room, the associated HVAC room, and stairway will maintain the fire barrier fire protection function as evaluated in the UFSAR Appendix 9A. The changes to the fire areas and fire zones do not affect the function of any safety-related structure, system, or component, and thus, do not introduce a new failure mode. The affected turbine building areas and equipment do not interface with any safety-related equipment or any equipment associated with radioactive material and, thus, do not create a new fault or sequence of events that could result in 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 proposed reconfiguration of the fire zones associated with the turbine building switchgear rooms, the electrical equipment room, and the associated HVAC room and the proposed editorial change will maintain the fire barrier fire protection function as evaluated in the UFSAR Appendix 9A. The fire barriers and equipment in the turbine building do not interface with any safety-related equipment or affect any safety-related function. The changes to the area barriers associated with the turbine building switchgear and associated HVAC continue to comply with the existing design codes and regulatory criteria, and do not affect any safety analysis.

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

NRC Branch Chief: Lawrence J. Burkhart.

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

Date of amendment request: July 24, 2014. A publicly-available version is in ADAMS under Accession No. ML14210A051.

Description of amendment request: The amendment would revise the reactor coolant pump (RCP) flywheel inspection surveillance requirements to extend the allowable inspection interval to 20 years, consistent with Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-421, “Revision to RCP Flywheel Inspection Program (WCAP-15666).” The Nuclear Regulatory Commission (NRC) staff believes that this amendment made use of both TSTF-421 and TSTF-237, Revision 1, “Relaxation of Reactor Coolant Pump Flywheel Examinations.”

The NRC staff published a notice of opportunity for comment in the Federal Register on June 24, 2003 (68 FR 37590), on possible amendments adopting TSTF-421, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated July 24, 2014.

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

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

The proposed change to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's acceptance guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiations exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences.

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

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

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the Start Printed Page 58828safety analysis assumptions and current plant operating practice.

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

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel.

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

The NRC staff has reviewed the 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: Jessie F. Quichocho.

Virginia Electric and Power Company, Docket Nos.: 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of amendment request: June 3, 2014. A publicly-available version is in ADAMS under Accession No. ML14160A607.

Description of amendment request: The amendments would revise the Surry Power Station (Surry) Units 1 and 2, Technical Specifications (TS). Specifically, TS Figures 3.1-1 and 3.1-2, Surry, Units 1 and 2, Reactor Coolant System Heatup Limitations and Surry, Units 1 and 2, Reactor Coolant System Cooldown Limitations, respectively, are being revised for clarification and to be fully representative of the allowable operating conditions during Reactor Coolant System startup and cooldown evolutions.

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 clarification of TS Figures 3.1-1 and 3.1-2 does not involve a physical change to the plant and does not change the manner in which plant systems or components are operated or controlled. The proposed change does not alter or prevent the ability of structures, system, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The P/T limits curves on TS Figures 3.1-1 and 3.1-2 are not being modified and remain valid.

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 clarification of TS Figures 3.1-1 and 3.1-2 does not involve any physical alteration of plant equipment; consequently, no new or different types of equipment will be installed. The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The P/T limits curves on TS Figures 3.1-1 and 3.1-2 are not being modified, and the basic operation of installed plant systems and components is unchanged.

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

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

Response: No.

The existing RCS P/T limits curves on TS Figures 3.1-1 and 3.1-2 are not being modified. The proposed clarification of TS Figures 3.1-1 and 3.1-2 does not alter any plant equipment, does not change the manner in which the plant is operated or controlled, and has no impact on any safety analysis assumptions. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The proposed change does not result in plant operation in a configuration outside the analyses or design basis and does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.

NRC Branch Chief: Robert J. Pascarelli.

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

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

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

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

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

Arizona Public Service Company, et al., Docket Nos.: STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona

Date of application for amendment: September 27, 2013, as supplemented by letter dated December 12, 2013.

Brief description of amendment: The amendments revised Technical Specification (TS) 3.3.3, “Control Element Assembly Calculators (CEACs),” to reinstate an inadvertently omitted 4-hour completion time for Start Printed Page 58829Required Action B.2.2. Additionally, the amendments revised a test frequency note in Surveillance Requirement (SR) 3.3.6.2 under TS 3.3.6, “Engineered Safety Features Actuation System (ESFAS) Logic and Manual Trip,” which should have been included in the license amendment request for Technical Specifications Task Force (TSTF) change traveler TSTF-425, Revision 3, “Relocate Surveillance Frequencies to Licensee Control—RITSTF [Risk-Informed TSTF] Initiative 5b.”

Date of issuance: September 9, 2014.

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

Amendment Nos.: Unit 1—194; Unit 2—194; Unit 3—194. A publicly-available version is in ADAMS under Accession No. ML14202A378; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendment revised the Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: February 4, 2014 (79 FR 6640). The supplemental letter dated December 12, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 9, 2014.

No significant hazards consideration comments received: No.

Dairyland Power Cooperative, Docket Nos.: 50-409 and 72-046, La Crosse Boiling Water Reactor (LACBWR), La Crosse County, Wisconsin

Date of application for amendment: August 6, 2013, supplemented by letters dated January 16, 2014, and April 14, 2014.

Brief description of amendment: The amendment approves changes to the Emergency Plan, including removal of the various emergency actions related to the former spent fuel pool, the transfer of responsibility for implementing the Emergency Plan to the Security Shift Supervisors at the ISFSI, a revised emergency plan organization, removal of the fire brigade, and abandonment of the LACBWR Control Room consistent with the current state of decommissioning, in that all of the spent nuclear fuel has now been transferred from the spent fuel pool to an independent spent fuel storage installation.

Date of issuance: September 8, 2014.

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

Amendment No.: 73.

Possession Only License No. DPR-45.

Date of initial notice in Federal Register: October 29, 2013 (78 FR 64543). The supplements dated January 16, 2014, and April 14, 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 on October 29, 2013 (78 FR 64543).

The Commission's related evaluation of the amendment is contained in a safety evaluation dated September 8, 2014.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos.: 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: October 28, 2013, as supplemented by letter dated June 3, 2014.

Brief description of amendments: The amendments modify Technical Specification (TS) 3.8.4. Specifically, the change allows a one-time extension of the completion time for Required Action A.2.2 to support replacement of the existing shared 125 VDC vital batteries.

Date of issuance: September 10, 2014.

Effective date: This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

Amendment Nos.: 274 and 254. A publicly-available version is in ADAMS under Accession No. ML14231A634; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and technical specifications.

Date of initial notice in Federal Register: January 21, 2014 (79 FR 3415). The supplemental letter dated June 3, 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 amendments is contained in a Safety Evaluation dated September 10, 2014.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit 2, Westchester County, New York

Date of amendment request: January 16, 2014, as supplemented by letters dated April 2, and April 15, 2014.

Brief description of amendment: The amendment revises Indian Point Nuclear Generating Unit 2 Technical Specification (TS) 5.5.7, “Steam Generator (SG) Program,” to exclude portions of the SG tubes below the top of the SG tubesheet from periodic inspections and plugging by implementing the alternate repair criteria “H*.” In addition, TS 3.4.13, “RCS [reactor coolant system] Operational Leakage,” is being revised to reduce the allowable primary to secondary leakage through any one SG from 150 to 85 gallons per day and TS 5.6.7, “Steam Generator Tube Inspection Program,” is being revised to include additional reporting requirements.

Date of issuance: September 5, 2014.

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

Amendment No.: 277. A publicly-available version is in ADAMS under Accession No. ML14198A161; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. DPR-26: The amendment revised the Facility Operating License and the TSs.

Date of initial notice in Federal Register: March 18, 2014 (79 FR 15147). The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 5, 2014.

No significant hazards consideration comments received: No.Start Printed Page 58830

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 application for amendment: May 21, 2013, as supplemented by letter dated October 4, 2013.

Brief description of amendment: The amendments revised the Technical Specifications (TSs) moderator temperature coefficient surveillance requirements associated with the implementation of Topical Report WCAP-16011-P-A, “Startup Test Activity Reduction (STAR) Program,” which describes the methods to be used for the implementation of reduction in the startup testing requirements. The changes are consistent with the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specifications change TSTF-486, Revision 2 as included in NUREG-1432, Revision 4.0, Standard Technical Specifications—Combustion Engineering Plants.

The NRC staff published a notice of opportunity for comment in the Federal Register on July 27, 2007 (72 FR 41360), on possible amendments adopting TSTF-486 using the NRC's consolidated line-item improvement process for amending licensees' TSs, which included a model safety evaluation (SE) and model no significant hazards consideration (NSHC) determination. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on September 6, 2007 (72 FR 51259), which included the resolution of public comments on the model SE and model NSHC determination. The licensee affirmed in its application dated May 21, 2013, that the proposed changes to the TSs satisfy the intent of TSTF-486.

Date of issuance: September 16, 2014.

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

Amendment Nos.: 219 and 168 (ADAMS Accession No. ML14218A180). Documents related to these amendments are provided in an SE enclosed with the amendments.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the License and TSs.

Date of initial notice in Federal Register: July 23, 2013 (78 FR 44173). The supplement dated October 4, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in an SE dated September 16, 2014.

No significant hazards consideration comments received: No.

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: November 21, 2013, as supplemented by letters dated February 4 and April 1, 2014.

Brief description of amendment: The amendments revised the date of cyber security plan (CSP) full implementation schedule (Milestone 8) and the existing license condition 2.H in the facility operating licenses NPF-87 and NPF-89 for CPNPP, Units 1 and 2, respectively. The CSP and the implementation schedule for CPNPP, Units 1 and 2, were previously approved by the NRC staff by letter dated July 26, 2011 (ADAMS Accession No. ML111780745).

Date of issuance: September 8, 2014.

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

Amendment Nos.: Unit 1—163; Unit 2—163. A publicly-available version is in ADAMS under Accession No. ML14183A342; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: April 8, 2014 (79 FR 19399). The supplements dated February 4 and April 1, 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 determination as published in the Federal Register.

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

No significant hazards consideration comments received: No.

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

Date of amendment request: January 30, 2012, as supplemented by letters dated May 10, 2012, September 20, 2012, March 27, 2013, December 20, 2013, and January 29, 2014.

Description of amendment request: The original application proposed revisions to the technical specifications (TSs) for new and spent fuel storage as a result of the new criticality analyses for the new fuel vault (NFV) and spent fuel pool (SFP). By letter dated December 20, 2013 (ADAMS Accession No. ML13360A045), NextEra requested that the SFP and NFV be separated into two separate license amendment requests. This amendment revised the TSs related to spent fuel storage as a result of new criticality analyses for the SFP. The license amendment request for the NFV will be processed under TAC No. MF3283.

Date of issuance: September 3, 2014.

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

Amendment No.: 142. A publicly-available version is in ADAMS under Accession No. ML14184A795; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-86: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: August 14, 2012 (77 FR 48559). The supplemental letters dated May 10, 2012, September 20, 2012, March 27, 2013, December 20, 2013, and January 29, 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.

Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 3, 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 4, 2014, as supplemented by the letter dated May 27, 2014.

Brief description of amendment: The amendment revises Tier 2* information, incorporated into the VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR). Specifically, the amendment revises the details regarding the structural floor of the Auxiliary Building and its constructability. Notes are added to drawings in subsection 3H.5 of the UFSAR in order to clarify variations in detail design such as size Start Printed Page 58831and spacing or reinforcement and spans of the noncritical sections of floors.

Date of issuance: July 3, 2014.

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

Amendment No.: 21. A publicly-available version is in ADAMS under Accession No. ML14150A133; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register: April 29, 2014 (79 FR 24025).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 3, 2014.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 22nd day of September 2014.

For the Nuclear Regulatory Commission.

George A. Wilson,

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

End Signature End Supplemental Information

[FR Doc. 2014-23015 Filed 9-29-14; 8:45 am]

BILLING CODE 7590-01-P