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Notice

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

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Start Preamble

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from June 12, to June 25, 2014. The last biweekly notice was published on June 24, 2014.

DATES:

Comments must be filed by August 7, 2014. A request for a hearing must be filed by September 8, 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-2014-0159. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

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FOR FURTHER INFORMATION CONTACT:

Sandra Figueroa, Office, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1262, email: Sandra.Figueroa@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-2014-0159 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-2014-0159.
  • 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-2014-0159 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 in comment submissions that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS, and 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 into ADAMS.

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

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

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

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license Start Printed Page 38586amendment 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.

B. Electronic Submissions (E-Filing)

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

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

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​getting-started.html. System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.Start Printed Page 38587

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

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

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

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having 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 obtaining information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: February 4, 2014. A publicly available version is in ADAMS under Accession No. ML14050A383.

Description of amendment request: The amendment would revise Technical Specification 5.5.15, “Containment Leakage Rate Testing Program,” to extend the frequency of the Type A, or the Containment Integrated Leak Rate Test, from 10 to 15 years on a permanent basis.

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

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

Response: No.

The proposed amendment involves changes to the IP3 [Indian Point Unit No. 3] containment leakage rate testing program. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The primary containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators.

Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

The proposed amendment adopts the NRC [Nuclear Regulatory Commission] accepted guidelines of [Nuclear Energy Institute] NEI 94-01, Revision 3-A, for development of the IP3 performance-based testing program for the Type A testing. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components would limit leakage rates to less than the Start Printed Page 38588values assumed in the plant safety analyses. The potential consequences of extending the ILRT [integrated leak rate test] interval to 15 years have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem per year within 50 miles resulting from design basis accidents was estimated to be acceptably small and determined to be within the guidelines published in [Regulatory Guide] RG 1.174. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. Entergy has determined that the increase in conditional containment failure probability due to the proposed change would be very small.

Therefore, it is concluded that the proposed amendment does not significantly increase the consequences of an 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 amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for the development of the IP3 performance-based leakage testing program, and establishes a 15-year interval for the performance of the containment ILRT. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

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

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

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for the development of the IP3 performance-based leakage testing program, and establishes a 15-year interval for the performance of the containment ILRT. This amendment does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the containment leakage rate testing program, as defined in the TS [technical specifications], ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant's safety analysis is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, Type B, and Type C containment leakage tests would be performed at the frequencies established in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 3-A.

Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that the containment would not degrade in a manner that is not detectable by an ILRT. A risk assessment using the current IP3 PSA [probabilistic safety assessment] model concluded that extending the ILRT test interval from ten years to 15 years results in a very small change to the risk profile.

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

NRC Branch Chief: Benjamin G. Beasley.

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

Date of amendment request: April 1, 2014. A publicly-available version is in ADAMS under Accession No. ML14099A227.

Description of amendment request: The amendments would revise the technical specifications by implementing Technical Specification Task Force Traveler 510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.”

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

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

Response: No.

The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. The proposed change to reporting requirements and clarifications of the existing requirements have no affect on the probability or consequences of a SGTR.

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

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

Response: No.

The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed changes do not affect the design of the SGs or their method of operation. In addition, the proposed changes do not impact any other plant system or component.

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

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

Response: No.

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

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

Changes associated with inspection frequency and tube selection criteria are consistent with TSTF-510 Revision 2 and are based on recent industry experience and are more effective in managing the frequency of verification of tube integrity and sample selection than those required by current TSs [technical specifications].

Therefore, the proposed change does not involve a significant reduction in any 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.Start Printed Page 38589

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

NRC Branch Chief: Benjamin G. Beasley.

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

Date of amendment request: October 8, 2013. A publicly-available version is in ADAMS under Accession No. ML13282A559.

Description of amendment request: The proposed amendment would revise the Technical Specification (TS) requirements to reduce the reactor pressure associated with the Reactor Core Safety Limit from 785 psig to 685 psig in TS 2.1.1.1 and TS 2.1.1.2. The proposed amendment would address the potential to not meet the lower pressure TS safety limit associated with a Pressure Regulator Failure-Maximum Demand (Open) (PRFO) transient reported by General Electric (GE) in their 10 CFR Part 21 Communication, Potential to Exceed Low Pressure Technical Specification Safety Limit, SC05-03, dated March 29, 2005.

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.

Decreasing the reactor pressure in TS Safety Limit 2.1.1.1 or 2.1.1.2 for reactor rated thermal power ranges effectively expands the validity range for GEXL correlation and the calculation of Minimum Critical Power Ratio Safety Limit (MCPR). The [critical power ratio] CPR rises during the pressure reduction following the scram that terminates the PRFO transient. Since the change does not involve a modification of any plant hardware, the probability and consequence of the PRFO transient are essentially unchanged. The reduction in the reactor dome pressure value in the safety limit from 800 psia (785 psig) to 700 psia (685 psig) provides greater margin to accommodate the pressure reduction during the transient within the revised TS limit.

The proposed change will continue to support the validity range for GEXL correlation and the calculation of MCPR as approved. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident or transient operating conditions.

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

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

Response: No.

The proposed reduction in the reactor pressure value in the safety limit from 800 psia (785 psig) to 700 psia (685 psig) reflects a wider range of applicability for the GEXL correlation for fuels in use at JAF and does not involve changes to the plant hardware or its operating characteristics. As a result, no new failure modes are being introduced.

Therefore, the change does not introduce a new or different kind of accident from those previously evaluated.

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

Response: No.

The margin of safety is established through the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. The proposed change in the reactor pressure safety limit enhances the safety margin, which protects the fuel cladding integrity during a depressurization transient, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The change does not alter the behavior of plant equipment, which remains unchanged. The available pressure range is expanded by the change, thus offering greater margin for pressure reduction during the transient.

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

Based on the above, Entergy concludes that the proposed amendment presents no 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: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Benjamin G. Beasley.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

Date of amendment request: January 29, 2014. A publicly-available version is in ADAMS under Accession No. ML14029A438.

Description of amendment request: The amendment would revise the facility operating license and technical specifications to reflect adoption of a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in NRC Regulatory Guide (RG) 1.205, Revision 1, “Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,” December 2009 (ADAMS Accession No. ML092730314). The license amendment request follows Nuclear Energy Institute (NEI) 04-02, Revision 2, “Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c),” April 2008. The submittal describes the methodology used to demonstrate compliance with, and transition to, National Fire Protection Association (NFPA) 805, and includes regulatory evaluations, probabilistic risk assessment, change evaluations, proposed modifications for non-compliances, and supporting attachments.

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:

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

Operation of Arkansas Nuclear One, Unit 1 (ANO-1) in accordance with the proposed amendment does not result in a significant increase in the probability or consequences of accidents previously evaluated. The proposed amendment does not affect accident initiators or precursors as described in the ANO-1 Safety Analysis Report (SAR), nor does it adversely alter design assumptions, conditions, or configurations of the facility, and it does not adversely impact the ability of structures, systems, or components (SSCs) to perform their intended function to mitigate the consequences of accidents described and evaluated in the SAR. The proposed changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions as required by the accident analysis. The SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition will remain capable of performing their design functions.

The purpose of this amendment is to permit ANO-1 to adopt a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance contained in Regulatory Guide (RG) 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection requirements that are an acceptable Start Printed Page 38590alternative to the 10 CFR Part 50, Appendix R, fire protection features (69 FR 33536; June 16, 2004).

The purpose of the fire protection program is to provide assurance, through defense-in-depth, that the NRC's fire protection objectives are satisfied. These objectives are: (1) preventing fires from starting; (2) rapidly detecting and controlling fires and promptly extinguishing those fires that do occur, thereby limiting fire damage; (3) providing an adequate level of fire protection for SSCs important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed; and (4) ensuring that fires will not significantly increase the risk of radioactive releases to the environment. In addition, fire protection systems must be designed such that their failure or inadvertent operation does not adversely impact the ability of the SSCs important to safety to perform their safety-related functions.

NFPA 805, taken as a whole, provides an acceptable alternative for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 10 CFR Part 50, meets the underlying intent of the NRC's existing fire protection regulations and guidance, and achieves defense-in-depth along with the goals, performance objectives, and performance criteria specified in NFPA 805, Chapter 1. In addition, if there are any increases in core damage frequency (CDF) or risk as a result of the transition to NFPA 805, the increase will be small, bounded by the delta risk requirements of NFPA 805, and consistent with the intent of the Commission's Safety Goal Policy.

Engineering analyses, which may include engineering evaluations, probabilistic risk assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been met. The SAR documents the analyses of design basis accidents (DBAs) at ANO-1. All accident analysis acceptance criteria will continue to be met with the proposed amendment. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions for the radiological consequence evaluations in the ANO-1 SAR. In addition, the applicable radiological dose acceptance criteria will continue to be met.

Based on the above, the implementation of this amendment to transition the Fire Protection Plan (FPP) at ANO-1 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a significant increase in the probability of any accident previously evaluated. In addition, all equipment required to mitigate an accident remains capable of performing the assumed function. Therefore, the consequences of any accident previously evaluated are not significantly increased with the implementation of this amendment.

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

Operation of ANO-1 in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Previously analyzed accidents with potential offsite dose consequences were included in the evaluation of the transition to NFPA 805. The proposed amendment does not impact these accident analyses. The proposed change does not alter the requirements or functions for systems required during accident conditions as assumed in the licensing basis analyses and/or DBA radiological consequences evaluations.

Implementation of the new risk-informed, performance-based fire protection licensing basis, which complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance contained in RG 1.205, will not result in new or different kinds of accidents. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004). No new modes of operation are introduced by the proposed amendment, nor will it create any failure mode not bounded by previously evaluated accidents. Further, the impacts of the proposed change are not directly assumed in any safety analysis to initiate an accident sequence.

The requirements in NFPA 805 address only fire protection and the impacts of fire effects on the plant have been evaluated. The proposed fire protection program changes do not involve new failure mechanisms or malfunctions that could initiate a new or different kind of accident beyond those already analyzed in the SAR. Based on this, as well as the discussion above, the implementation of this amendment to transition the FPP at ANO-1 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), 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

Operation of ANO-1 in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. The transition to a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) 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 affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed in the SAR to mitigate accidents. The proposed change does not adversely impact systems that respond to safely shut down the plant and maintain the plant in a safe shutdown condition. In addition, the proposed amendment will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without implementation of appropriate compensatory measures.

The risk evaluations for plant changes, in part as they relate to the potential for reducing a safety margin, were measured quantitatively for acceptability using the delta risk (i.e., ΔCDF and ΔLERF [large early release frequency]) criteria from Section 5.3.5, “Acceptance Criteria,” of NEI 04-02, as well as the guidance contained in RG 1.205. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based methods of NFPA 805 do not result in a significant reduction in the margin of safety. As such, the proposed changes are evaluated to ensure that risk and safety margins are kept within acceptable limits. Based on the above, the implementation of this amendment to transition the FPP at ANO-1 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), will not significantly reduce 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: Joseph A. Aluise, Associate General Council—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, LA 70113.

NRC Branch Chief: Michael T. Markley.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: April 30, 2014. A publicly-available version is in ADAMS under Accession No. ML14127A435.

Description of amendment request: The proposed amendment would revise Oyster Creek Nuclear Generating Station (OCNGS) Technical Specification (TS) 4.5 M., “Shock Suppressors (Snubbers),” to conform the TS to the revised OCNGS Snubber Inspection Program.

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

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

Response: No.

The proposed changes would revise TS 4.5.M to conform the TS to the revised Snubber Inspection Program. Snubber examination, testing and service life Start Printed Page 38591monitoring will continue to meet the requirements of 10 CFR 50.55a(g). Snubber examination, testing and service life monitoring is not an initiator of any accident previously evaluated.

Therefore, the probability of an accident previously evaluated is not significantly increased.

Snubbers will continue to be demonstrated OPERABLE by performance of a program for examination, testing and service life monitoring in compliance with 10 CFR 50.55a or authorized alternatives. The proposed changes do 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.

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

2. Does the amendment 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 these proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes 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.

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

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

Attorney for licensee: J. Bradley Fewell, VP & Deputy General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Meena Khanna.

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

Date of amendment request: April 9, 2014. A publicly-available version is in ADAMS under Accession No. ML14101A367.

Description of amendment request: The proposed amendment would revise the Donald C. Cook Nuclear Plant, Units 1 and 2, technical specification (TS) 3.4.2, “[Reactor Coolant System (RCS)] Pressure and Temperature (P/T) Limits,” to address an issue regarding the applicability of TS Figures 3.4.3-1 “Reactor Coolant System Pressure versus Temperature Limits—Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 [Effective Full Power Years (EPFY)]” and 3.4.3-2 “Reactor Coolant System Pressure versus Temperature Limits—Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY)” during vacuum fill operations of the RCS.

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 TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. There are no physical changes to the plant being introduced by the proposed changes to the heatup and cooldown limitation curves. The proposed changes do not modify the RCS pressure boundary. That is, there are no changes in operating pressure, materials, or seismic loading. The proposed changes do not adversely affect the integrity of the RCS pressure boundary such that its function in the control of radiological consequences is affected.

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

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

Response: No.

The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents. Further, the proposed changes to the heatup and cooldown limitation curves do not affect any activities or equipment other than the RCS pressure boundary and do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

The proposed TS changes do not involve a significant reduction in the margin of safety. The revised heatup and cooldown limitation curves and low-temperature overpressure protection limits are established in accordance with current regulations and the [American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV)] Code 1995 edition with 1996 Addenda. These proposed changes are acceptable because the ASME B&PV Code maintains the margin of safety required by [Title 10 of the Code of Federal Regulations (10 CFR)] 50.55(a). Because operation will be within these limits, the RCS materials will continue to behave in a non-brittle manner consistent with the original design bases.

Therefore, the proposed amendment does not involve a significant reduction in 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: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: October 4, 2013, as supplemented by letter dated April 29, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML13281A826 and ML14122A044, respectively.

Description of amendment request: Following completion of an on-site staffing analysis of the Emergency Response Organization, NSPM determined that the Radwaste Operator is no longer required to augment plant staff for performing repairs and corrective actions as prescribed in the MNGP Emergency Plan. The amendment proposes to remove the Radwaste Operator position as a 60-minute responder credited within the MNGP Emergency Plan.

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 38592issue 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 to the Emergency Plan does not impact the function of plant structures, systems, or components (SSCs). The proposed change does not affect accident initiators or precursors, nor does it alter design assumptions. The proposed change does not alter or prevent the ability of the Emergency Response Organization to perform their intended functions to mitigate the consequences of an accident or event. This proposed change only removes a no longer credited position from the Emergency Plan.

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

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

Response: No.

The proposed change does not impact the accident analysis. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The proposed change does not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. This proposed change only removes a no longer credited position from the Emergency Plan. The proposed change, therefore, does not alter or prevent the ability of the Emergency Response Organization to perform their intended functions to mitigate the consequences of an accident or event.

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.

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 change is associated with the Emergency Plan staffing and does not impact operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed change does not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed change. Safety analysis acceptance criteria are not affected by this proposed change. The revised Emergency Plan will continue to provide the necessary response staff with the proposed 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: Robert D. Carlson.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: February 10, 2014, as supplemented by letter dated June 9, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14041A408 and ML14163A417, respectively.

Description of amendment request: The proposed amendment would revise the Technical Specification (TS) surveillance frequency for the pressurizer safety valves from a refueling frequency (i.e., 18 months +25 percent) to be consistent with the Inservice Testing Program. In addition, the proposed amendment would administratively change the format of the footnotes in TS Table 3-5, “Minimum Frequencies for Equipment Tests.”

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 requested change revises the performance interval of one TS surveillance requirement to be consistent with the Inservice Testing Program as stated in 10 CFR 50.55a(g)(5). The performance of the surveillance, or the failure to perform the surveillance, is not a precursor to an accident. Performing the surveillance or failing to perform the surveillance does not affect the probability of an accident. Even with the requested extension, the period during which the plant is in Modes 1 or 2 and the valves are required to be operable will be no longer than a typical operating cycle. Also, the proposed interval between tests will be consistent with the interval for this type of valve specified by the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 1998 Edition, through 2000 Addenda, Appendix I, frequency requirements for testing of pressure relief valves.

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

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

Response: No.

The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Hence, the proposed change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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

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

Response: No.

The proposed change revises the performance interval for one surveillance requirement to be consistent with the test interval for this type of valve specified by the ASME OM Code, 1998 Edition, through 2000 Addenda as required by 10 CFR 50.55a. This change does not alter any safety margins.

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

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

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

NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: March 31, 2014. A publicly-available version is in ADAMS under Accession No. ML14090A417.

Description of amendment request: The proposed amendment would change Technical Specification 2.5, Auxiliary Feedwater (AFW) system to allow a 7-day completion time for the turbine-driven AFW pump if the inoperability occurs following a refueling outage and if MODE 2 had not been entered.

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 Start Printed Page 38593consequences of an accident previously evaluated?

Response: No.

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

The proposed change to Technical Specification (TS) 2.5 would allow a seven day Completion Time for the turbine-driven Auxiliary Feedwater (AFW) pump if the inoperability occurs following a refueling outage, and if MODE 2 had not been entered. The note currently in TS 2.5 Applicability addresses the issue of allowing additional time to perform necessary testing to prove the operability of the turbine driven AFW pump following refueling as approved by the NRC in TS Amendment 127. This note does not specifically state that it is only allowed following refueling and does not restrict the time the plant can be in this condition. The proposed change will be more restrictive than the current TS since it will specifically state when it is allowed (following refueling) and for how long it is allowed.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because: 1) the proposed amendment does not represent a change to the system design, 2) the proposed amendment does not prevent the safety function of the AFW system from being performed, since the other fully redundant essential train is required to be operable, 3) the proposed amendment does not alter, degrade, or prevent action described or assumed in any accident Updated Safety Analysis Report (USAR) from being performed since the other train of AFW is required to be operable, 4) the proposed amendment does not alter any assumptions previously made in evaluating radiological consequences, and 5) the proposed amendment does not affect the integrity of any fission product barrier. No other safety related equipment is affected by the proposed change.

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

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

Response: No.

The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Hence, the proposed change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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

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

Response: No.

The proposed changes do not involve a significant reduction in a margin of safety. The proposed change to TS 2.5 would restrict for the turbine-driven AFW pump inoperability to a seven day Completion Time if the inoperability occurs following a refueling outage and prior to MODE 2 being entered. The current Note in TS 2.5 Applicability does not require the turbine driven AFW pump to be operable until prior to entering MODE 2; therefore, the proposed change is more restrictive than current TS.

The proposed change does not involve a significant reduction in a margin of safety because: (1) during a return to power operations following a refueling outage, decay heat is at its lowest levels, (2) the other AFW train is required to be operable, and (3) the motor-driven AFW train can provide sufficient flow to remove decay heat and cool the unit to shutdown cooling system entry conditions from power operations. This change does not alter any safety margins.

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

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

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

NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: April 25, 2014. A publicly-available version is in ADAMS under Accession No. ML14118A435.

Description of amendment request: The proposed amendment would revise Section 5.11, “Structures Other Than Containment,” and Appendix F, “Classification of Structures and Equipment and Seismic Criteria,” of the Fort Calhoun Station, Unit No. 1, Updated Safety Analysis Report. The changes would clarify the licensing and design basis to permit the use of seismic floor response spectra in analysis and design of seismic Class I structures and structural elements attached to structures.

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

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

Response: No.

[T]his change to the Updated Safety Analysis Report (USAR) has no effect on the consequences of any accident, as it makes no physical changes to the plant. Since the Alternate Seismic Criteria and Methodologies (ASCM) floor response spectra (FRS) represent a refined version of the plant's original design basis, the design margins for any application utilizing the FRS will be maintained with respect to the design basis earthquake. Thus, the proposed amendment does not result in 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.

[T]he change to the USAR does not change any accident analyses, does not make any physical changes to the plant, and does not change the way the plant is operated. The only change is to permit the utilization of the ASCM curves in the design and evaluation of structural applications. The curves themselves are based on the same earthquake as the plant's original design. Thus, 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.

[T]he ASCM FRS is based on the same earthquake as the plant's original design basis. The ASCM FRS are refined curves of the same design basis and thus, the design margins of any application or evaluation utilizing the ASCM FRS will be maintained with respect to the design basis earthquake. Thus, 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 A. Repka, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 16, 2014. A publicly-available version is in ADAMS under Accession No. ML14143A370.

Description of amendment request: The proposed amendment would revise the Updated Safety Analysis Report (USAR) to allow pipe stress analysis of non-reactor coolant system safety-related piping to be performed in accordance with the American Society Start Printed Page 38594of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section III, 1980 Edition as an alternative to current Code of record.

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

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

Response: No.

The proposed change to the current licensing basis (CLB) allows the use of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section III, 1980 Edition (no Addenda) as an alternative to the original Code of Record (i.e., United States of America Standards (USAS) B31.7 1968 (DRAFT) Edition) for the design and analysis of non-reactor coolant system (RCS) piping. The American National Standards Institute (ANSI) B31 Code Committee has determined that:

“. . . piping that has been designed and constructed in accordance with Section III of the ASME Boiler and Pressure Vessel Code including addenda and applicable cases may be accepted as complying with the requirements of B31.7, 1969 and applicable addenda for the respective class of construction.”

Although the ANSI B31 Code Committee statement refers to the B31.7, 1969 Edition, there are no significant differences between it and the B31.7 1968 (DRAFT) Edition. The change involves the substitution of one accepted piping Code for another and not a physical plant change. The Updated Safety Analysis Report (USAR) accident analysis assumes the proper functioning of safety systems in demonstrating the adequacy of the plant's design. This change does not alter the intended function of any plant equipment nor does it degrade or increase challenges to the performance of safety systems assumed to function in the accident analysis.

The use of ASME BPV Code, Section III, 1980 Edition (no Addenda) analytical methods provides acceptable design results with no reduction in radiological barrier safety margin. Hence, there is no change in radiological barrier performance that would increase the dose to personnel onsite (10 CFR 20) or to the public at the site boundary (10 CFR 100).

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

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

Response: No.

The proposed amendment provides the basis for the use of ASME BPV Code, Section III, 1980 Edition (no Addenda) for stress analysis of non-RCS safety-related piping. This approach will not introduce any methods or analytical techniques that could create the possibility of a new or different kind of accident. Application of a Code methodology does not create the possibility of a different kind of accident.

The application of the ASME BPV Code, Section III, 1980 Edition (no Addenda) does not create any new unanalyzed interactions between systems or components. Piping systems will be analyzed in accordance with the Code, which is one part of the framework to establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety. The proposed change to the CLB does not create a new failure mechanism or new accident initiator. The proposed amendment does not involve a change in methods governing the operation of plant systems or components.

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

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

Response: No.

The Fort Calhoun Station Technical Specifications (TS) ensure that the plant operates in a manner that will ensure acceptable levels of protection for the health and safety of the public. The Technical Specifications ensure that the available equipment and initial conditions for a Design Basis Accident (DBA) as defined in the USAR meet the assumptions in the accident analysis contained in the USAR. The plant safety margins are addressed in the Technical Specification Bases and the USAR.

This proposed amendment revises the CLB to allow the use of ASME BPV Code, Section III, 1980 Edition (no Addenda) for stress analysis of non-RCS safety-related piping. No changes are being made to the physical plant. The use of the ASME BPV Code, Section III, 1980 Edition (no Addenda) does not change, revise, or otherwise affect the current Technical Specifications (TS) or TS Bases. Incorporation of the ASME BPV Code, Section III, 1980 Edition (no Addenda) into the FCS CLB will not affect the current plant design parameters or TS Limiting Conditions for Operation (LCO).

The proposed change does not modify, change, revise, or otherwise affect any current calculations concerning the plant accident analysis or supporting basis for which the TSs, TS Bases, or USAR safety margins were established. 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 A. Repka, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Michael T. Markley.

ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station (ZNPS), Units 1 and 2, Lake County, Illinois

Date of amendment request: March 17, 2014. A publicly-available version is in ADAMS under Accession No. ML14078A049.

Description of amendment request: The proposed amendments would amend licenses DPR-39 and DPR-48 and revise the Zion Technical Specifications (TS) to reflect the removal of all the spent fuel from the Zion spent fuel pool. The proposed changes to both Facility Operating Licenses modify Section 2.C.(6) to specify the ZNPS Independent Spent Fuel Storage Installation Physical Security Plan (ISFSI), eliminate Section 2.C.(7) Spent Fuel Pool Modification, and eliminate Section 2.C.(16), related to the single-failure proof fuel building crane. The proposed changes to the TS eliminate provisions of the specifications applicable to spent fuel stored in the spent fuel pool and relocate the remaining TS administrative requirements to the Quality Assurance Project Plan. These changes are proposed pursuant to the criteria contained in 10 CFR 50.36 and in accordance with the recommendations contained in the U.S. Nuclear Regulatory Commission's (NRC) Administrative Letter 95-06. The proposed changes will result in a TS that will be applicable to the ZNPS once the last spent fuel assembly has been removed from the spent fuel pool and placed at the ISFSI.

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

The proposed changes (deletion of operational requirements and certain design requirements) reflect the complete transfer of the spent fuel from the spent fuel pool to the ISFSI. Design basis accidents related to the spent fuel pool are discussed in the ZNPS Defueled Safety Analysis Report (DSAR) Chapter 5. These postulated accidents are predicated on spent fuel being stored in the spent fuel pool. With the removal of the spent fuel from the spent fuel pool, there are no remaining spent fuel assemblies to be monitored and there are no credible accidents that require the actions of a Certified Fuel Handler, Shift Supervisor, or a Non-certified Operator to prevent occurrence or mitigate the consequences of an accident.

In addition, the ZNPS DSAR Chapter 5 also provides analyses of accidents as result of Start Printed Page 38595decommissioning with the bounding consequences resulting from the failure of a High Integrity Container (HIC) containing dewatered radioactive demineralizer resin.

The proposed changes do not have an adverse impact on the remaining decommissioning activities or any decommissioning related postulated accident consequences.

The proposed changes related to the relocation of certain administrative requirements do not affect operating procedures or administrative controls that have the function of preventing or mitigating any remaining decommissioning design basis accidents. In addition, these proposed changes are consistent with the guidance of the NRC's Administrative Letter 95-06.

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

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

The proposed changes eliminate the operational requirements and certain design requirements associated with the storage of the spent fuel in the spent fuel pool, and relocate certain administrative controls to the Quality Assurance Program Plan.

With the complete removal of the spent fuel from the spent fuel pool and transfer to the ISFSI, there are no spent fuel assemblies that remain at the plant and the potential for fuel related accidents is removed. The proposed changes do not introduce any new failure modes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Does the change involve a significant reduction in a margin of safety?

The design basis and accident assumptions within the ZNPS DSAR and the TS relating to spent fuel are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities. In addition, the proposed changes do not result in a change in initial conditions, system response time, or in any other parameter affecting the course of the remaining decommissioning activity accident analysis. 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: Russ Workman, Deputy General Counsel, EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 84101.

NRC Branch Chief: Bruce Watson.

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 “Accessing 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, Unit Nos. 1, 2, and 3, Maricopa County, Arizona

Date of application for amendment: December 26, 2012, as supplemented by letter dated August 26, 2013.

Brief description of amendment: The amendments adopt Technical Specifications Task Force (TSTF) change traveler TSTF-500, Revision 2, “DC Electrical Rewrite—Update to TSTF-360.” The amendments revised TS requirements related to direct current (DC) electrical systems in TS limiting condition for operation (LCO) 3.8.4, “DC Sources—Operating,” LCO 3.8.5, “DC Sources—Shutdown,” and LCO 3.8.6, “Battery Parameters.” A new “Battery Monitoring and Maintenance Program” was added to Section 5.5, “Programs and Manuals.”

Date of issuance: June 25, 2014.

Effective date: As of the date of issuance and shall be implemented within 180 days from the date of issuance.

Amendment No.: Unit 1—193; Unit 2—193; Unit 3—193. A publicly-available version is in ADAMS under Accession No. ML14115A045; 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 amendments revised the Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: March 4, 2013 (78 FR 14129). The supplement dated August 26, 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 June 25, 2014.

No significant hazards consideration comments received: No.

Dominion Energy Kewaunee, Inc. (DEK), Docket No. 50-305, Kewaunee Power Station (KPS), Kewaunee County, Wisconsin

Date of application for amendment request: April 16, 2013, as supplemented by letters dated September 5, 2013, October 14, 2013, and March 19, 2014.

Brief description of amendment: The amendment revised the Renewed Facility Operating License by deleting a license condition associated with license renewal and adding a license condition related to spent fuel pool storage rack boron absorber surveillance.

Date of issuance: June 23, 2014.

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

Amendment No.: 213. A publicly-available version is in ADAMS under Accession No. ML14008A297; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-43: Amendment revised the Renewed Facility Operating License.Start Printed Page 38596

Date of initial notice in Federal Register: August 20, 2013 (78 FR 51223). The supplemental letters dated September 5, 2013, October 14, 2013, and March 19, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 23, 2014.

No significant hazards consideration comments received: No.

Dominion Energy Kewaunee, Inc. (DEK), Docket No. 50-305, Kewaunee Power Station (KPS), Kewaunee County, Wisconsin

Date of application for amendment request: May 29, 2013, as supplemented by letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and letters dated January 7, 2014, and March 13, 2014.

Brief description of amendment: The amendment revised the Renewed Facility Operating License Technical Specifications (TSs) to permit fuel handling activities consistent with the permanently shutdown and defueled condition of the facility. Specifically, in its March 13, 2014, supplemental letter DEK stated that it had accelerated the schedule to transfer spent fuel from the spent fuel pool to the independent spent fuel storage installation (ISFSI). Under its new schedule, DEK plans to begin activities to support spent fuel transfer to the ISFSI by July 1, 2014. Based on its new schedule, DEK requested expedited review and partial approval of the deletion of certain TSs currently required for movement of irradiated fuel assemblies. If not amended, the affected TSs would require restoring operability of certain equipment during spent fuel handling activities that are no longer needed for accident mitigation.

The NRC staff has issued a partial approval of the original May 29, 2013, amendment request as supplemented, to permit fuel handling activities in accordance with DEK's request in its March 13, 2014, submittal. The staff continues to review the remaining license condition and technical specification changes requested in DEK's May 29, 2013, submittal as supplemented, that were not addressed in this amendment.

Date of issuance: June 9, 2014.

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

Amendment No.: 212. A publicly-available version is in ADAMS under Accession No. ML14111A234; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-43: Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: August 20, 2013 (78 FR 51224). The supplemental letters dated September 23, October 15, October 17, October 31, and November 7, 2013, January 7, 2014, and March 13, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

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

No significant hazards consideration comments received: No.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: February 7, 2013, as supplemented by letter dated January 16, 2014.

Brief description of amendment: The amendment revised the River Bend Station, Unit 1 (RBS) Technical Specification (TS) 3.8.4, “DC [Direct Current] Sources—Operating,” Surveillance Requirements 3.8.4.2 and 3.8.4.5. The change is the result of the licensee's determination that the total battery capacity would possibly be insufficient to supply the required load to the DC system if each of the battery-to-battery connections were to reach the individual resistance limits. The changes to the Surveillance Requirements added new acceptance criteria to address the possible non-conservative conditions when the battery connection resistances are at maximum TS values.

Date of issuance: June 18, 2014.

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

Amendment No.: 181. A publicly-available version is in ADAMS under Accession No. ML14136A008; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-47: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: April 30, 2013 (78 FR 25312). The supplemental letter dated January 16, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 18, 2014.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 14, 2012, as supplemented by letters dated January 29, February 14, May 30, and October 22, 2013, and March 11, 2014.

Brief description of amendment: The amendments revised the operating licenses and Technical Specifications (TSs) to remove completed and satisfied license conditions, revised TS 5.5.1 to remove related conditions, corrected inadvertent errors, updated references to the Physical Security Plan, and made editorial changes to the operating licenses and TSs.

Date of issuance: June 13, 2014.

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

Amendment Nos.: 260 and 255. A publicly-available version is in ADAMS under Accession No. ML13329A092; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the licenses and the TSs.

Date of initial notice in Federal Register: January 8, 2013 (78 FR 1271), and April 16, 2013 78 FR 22569). The submittal dated January 29, 2013, expanded the scope of the application dated September 14, 2012, and the application was renoticed April 16, 2013. The supplements dated February 14, May 30, and October 22, 2013, and March 11, 2014, provided additional information that clarified the application, did not expand the scope of the submittal dated January 29, 2013, as noticed, and did not change the staff's proposed no significant hazards consideration determinations published on January 8, 2013, and April 16, 2013. The supplement dated March 11, 2014, limited the scope of the supplement dated January 29, 2013, by deleting the Start Printed Page 38597proposed change to TS Figure 3.1-2, “Boric Acid Tank Minimum Volume.”

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 13, 2014.

No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of application for amendment: September 28, 2011, as supplemented by letters dated December 19 and December 22, 2011; March 20, July 24, August 24, and September 27, 2012; April 23, May 21, July 29, September 12, October 11, November 4, November 11, and December 18, 2013; and January 24, February 28, April 10, and June 11, 2014.

Brief description of amendment: The amendment transitions the Fort Calhoun Station fire protection program to a risk-informed, performance-based program based on National Fire Protection Association (NFPA) 805, in accordance with 10 CFR 50.48(c). NFPA 805 allows the use of performance-based methods such as fire modeling and risk-informed methods such as fire probabilistic risk assessment to demonstrate compliance with the nuclear safety performance criteria.

Date of issuance: June 16, 2014.

Effective date: As of its date of issuance and shall be implemented by 12 months from the date of issuance.

Amendment No.: 275. A publicly-available version is in ADAMS under Accession No. ML14098A092; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-40: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: April 10, 2012 (77 FR 21598). The supplements dated March 20, July 24, August 24, and September 27, 2012; April 23, May 21, July 29, September 12, October 11, November 4, November 11, and December 18, 2013; and January 24, February 28, April 10, and June 11, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 16, 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: December 20, 2013.

Brief description of amendment: The amendment revises the plant's emergency plan. In conjunction with the new license condition, the amendment complies with the established regulatory changes set forth in “Enhancements to Emergency Preparedness Regulations,” published in the Federal Register on November 23, 2011 (76 FR 72560). Specifically, the license amendment changes on-shift staffing analysis and the changes to the emergency plan address evacuation time estimates. The design, construction and operation of the plant are not affected by this license amendment and license condition.

Date of issuance: May 30, 2014.

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

Amendment No.: 20. A publicly-available version is in ADAMS under Accession No. ML14118A252; 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: February 4, 2014, (79 FR 6643).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 30, 2014.

No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

Date of application for amendment: December 13, 2012, as supplemented by letters dated June 11, 2013, and January 16 and April 9, 2014.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.7.9, “Ultimate Heat Sink (UHS),” to incorporate more restrictive UHS level and pond temperature limits which are specified in Surveillance Requirements (SRs) 3.7.9.1 and 3.7.9.2, respectively. In addition, new SR 3.7.9.4 is added to verify that the UHS cooling tower fans respond appropriately to automatic start signals.

Date of issuance: June 17, 2014.

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

Amendment No.: 208. A publicly-available version is in ADAMS under Accession No. ML14149A164; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: March 4, 2013 (78 FR 14138). The supplements dated June 11, 2013, and January 16 and April 9, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 17, 2014.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 27th day of June, 2014.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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

End Signature End Supplemental Information

[FR Doc. 2014-15770 Filed 7-7-14; 8:45 am]

BILLING CODE 7590-01-P