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Notice

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

Document Details

Information about this document as published in the Federal Register.

Published Document

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

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Background

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 January 22, 2014 to February 5, 2014. The last biweekly notice was published on January 21, 2014 (79 FR 3412).

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-2014-0028. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: 3WFN-06-44M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

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SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2014-0028 when contacting the NRC about the availability of information regarding this document. You may access publicly-available information related to this document by any of the following methods:

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0028.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library 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 in this document (if that document is available in ADAMS) is provided the first time that a document is referenced.
  • 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-2014-0028 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

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

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

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

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 Start Printed Page 9491within 30 days after the date of publication of this notice will be considered in making any final determination.

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

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-rmdoc-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 reques to petitioner to relief. A request or 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.

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

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

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.html. System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the NRC'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 Start Printed Page 9492unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

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

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with the NRC 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 agency'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 Web site at http://www.nrc.gov/​site-help/​e-submittals.html, by email at 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 this license amendment application, see the application for amendment which is available for public inspection at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov.

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

Date of amendment request: September 26, 2013.

Description of amendment request: The amendments requests transition of the fire protection licensing basis at MNS, Units 1 and 2, from §§ 50.48(b) and 50.48(c) of Title 10 of the Code of Federal Regulations (10 CFR), National Fire Protection Association (NFPA) 805.

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

Response: No.

Operation of MNS in accordance with the proposed amendment does not increase the probability or consequences of accidents previously evaluated. The Updated Final Safety Analysis Report documents the analyses of design basis accidents at MNS. The proposed amendment does not adversely affect accident initiators nor alter design assumptions, conditions, or configurations of the facility and does not adversely affect the ability of structures, systems, and components to perform their design function. Structures, systems, and components required to safely shut down the reactor and to maintain it in a safe shutdown condition will remain capable of performing their design functions.

One purpose of this amendment is to permit MNS to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and Start Printed Page 9493the guidance 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 system and features that are an acceptable alternative to the Appendix R fire protection features (69 FR 33536; June 16, 2004). Engineering Analyses, in accordance with NFPA 805, have been performed to demonstrate that the risk-informed performance-based requirements for NFPA 805 have been met.

The NFPA 805, taken as a whole, provides an acceptable alternative to 10 CFR 50.48(b) and satisfies 10 CFR 50.48(a) and General Design Criterion 3 of Appendix A to 10 CFR Part 50 and meets the underlying intent of the NRC's existing fire protection regulations and guidance, and achieves defense-in-depth and the goals, performance objectives, and performance criteria specified in Chapter 1 of the standard. The increases in core damage frequency associated with the LAR submittal are acceptable within the guidance of RG 1.174, therefore this allows self approval of the fire protection program changes post-transition. If there are any increases post-transition in core damage frequency or risk, the increase will be small and consistent with the intent of the Commission's Safety Goal Policy.

Based on this, the implementation of this proposed amendment does not significantly increase the probability of any accident previously evaluated. 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 the amendment.

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

Response: No.

Operation of MNS in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Any scenario or previously analyzed accident with offsite dose was included in the evaluation of design basis accidents documented in the Updated Final Safety Analysis Report. The proposed change does not alter the requirements or function for systems required during accident conditions. Implementation of the new Fire Protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in RG 1.205 will not result in new or different accidents.

The proposed amendment does not adversely affect accident initiators nor alter design assumptions, conditions, or configurations of the facility. The proposed amendment does not adversely affect the ability of structure, systems, and components to perform their design function. Structure, systems, and components required to safely shut down the reactor and maintain it in a safe shutdown condition remain capable of performing their design functions.

The purpose of this amendment is to permit MNS to adopt a new Fire Protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in RG 1.205. 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 Appendix R Fire Protection features (69 FR 33536; June 16, 2004).

The requirements in NFPA 805 address only Fire Protection and the impacts of fire on the plant have already been evaluated. Based on this, the implementation of this proposed amendment does not create the possibility of a new or different kind of accident from any kind of accident previously evaluated. The proposed changes do not involve new failure mechanisms or malfunctions that can initiate a new accident.

Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created with the implementation of this amendment.

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

Response: No.

Operation of MNS in accordance with the proposed amendment does not involve a significant reduction in the margin of safety. The proposed amendment 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 to mitigate accidents in the Updated Final Safety Analysis Report. The proposed amendment does not adversely affect the ability of Structure, Systems, and Components to perform their design function. Structure, Systems, and Components required to safely shut down the reactor and to maintain it in a safe shutdown condition remain capable of performing their design functions.

One purpose of this amendment is to permit MNS to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in RG 1.205. 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 McGuire Nuclear Station's existing fire protection requirements. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based methods do not result in a significant reduction in the margin of safety.

Based on this, the implementation of this proposed amendment does not significantly reduce the margin of safety. The proposed changes are evaluated to ensure that risk and safety margins are kept within acceptable limits. Therefore, the transition does not involve a significant reduction in the margin of safety.

The NFPA 805 continues to protect public health and safety because the overall approach of NFPA 805 is consistent with the key principles for evaluating license basis changes, as described in RG 1.174, is consistent with the defense-in-depth philosophy, and maintains sufficient safety margins.

Margins previously established for the MNS Fire Protection program in accordance with existing fire protection requirements are not significantly reduced.

Therefore, this proposed amendment does not result in a reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the 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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station (ONS), Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: October 24, 2013.

Description of amendment request: The proposed amendments would revise Section 3.1.1.1 of the Updated Final Safety Analysis Report (UFSAR) for ONS Units 1, 2, and 3 to clarify quality requirements of the Standby Shutdown Facility (SSF) and interconnected systems.

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 involves no change to the plant design and is intended to ensure a consistent interpretation of wording previously included in the UFSAR regarding the QA classification of certain Structures, Systems, and Components (SSCs) relied upon to address a postulated Turbine Building flood event. The proposed change will help to ensure the design of the SSF is maintained consistent with the licensed design. The proposed UFSAR change does not involve operating any installed equipment in a new or different manner or a change to any set points for parameters which initiate protective or mitigation action. There is no adverse impact on containment integrity, radiological release pathways, fuel design, filtration systems, main steam relief valve set Start Printed Page 9494points, or radwaste systems. No new radiological release pathways are created. Because this correction and clarification to the UFSAR design description does not alter the SSF design as licensed, the proposed change does not involve a significant increase in the probability or consequences of any event requiring operation of the SSF.

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 requests approval to modify and clarify a UFSAR design description to ensure the described design of the ONS units and the SSF is maintained consistent with the licensed design. In accordance with this revision, replacement equipment is functionally equivalent to the existing and is designed to the appropriate pressure, temperature, and environmental parameters. The proposed change does not change the design function or operation of the SSF or of the interconnecting seismic induced turbine building flood equipment. Further, the proposed change does not create a new or different kind of accident since the proposed changes do not introduce credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

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

Response: No.

The proposed change requests approval to modify and clarify a UFSAR design description to ensure a consistent understanding of the licensed design of the plant, including the SSF. The proposed change does not change the design function or operation of the SSF. The proposed change does not involve operating any installed equipment in a new or different manner; a change to any set points for parameters which initiate protective or mitigation action; or any impact on the fission product barriers or safety limits.

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: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202-1802.

NRC Branch Chief: Robert J. Pascarelli.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: October 31, 2013.

Description of amendment request: The licensee has indicated their intent to submit certifications pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.82(a)(1)(i) and (ii) along with 10 CFR 50.82(a)(2) committing to the permanent cessation of operations and the permanent removal of fuel from the reactor vessel. Following these certifications, the 10 CFR part 50 operating license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. The proposed amendment includes a number of changes to revise or eliminate current requirements found in Section 6.0, Administrative Controls, of the Vermont Yankee Technical Specifications to support a defueled reactor, the new organization, and the permanent shutdown of the facility. Proposed changes include (1) elimination of the Mitigating Strategies License Condition in the operating license, (2) revisions to Section 6.1, Responsibility, regarding control room command function and delegation of authority, (3) revisions to Section 6.2, Organization, to reflect emphasis on the safe handling and storage of spent nuclear fuel as opposed to nuclear plant operations along with the conversion of license reactor operators to certified fuel handlers, (4) elimination of Section 6.3, Actions to be Taken if a Safety Limit is Exceeded, (5) revision to Section 6.4, Procedures, to reflect a permanently defueled reactor vessel, (6) revision to Section 6.6, Reporting Requirements, to eliminate the Core Operating Limits Report, and (7) revision to Section 6.7, Programs and Manuals to eliminate the Integrity of Systems Outside Containment program, eliminate the Plant Offsite Review Committee review of changes to the Offsite Dose Calculation Manual, and eliminate the Primary Containment Leakage Rate Testing 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 examined?

Response: No.

The proposed amendment would not take effect until VY [Vermont Yankee Nuclear Power Station] has permanently ceased operation and entered a permanently defueled condition. The proposed amendment would modify the VY OL [operating license] and TS [technical specifications] by deleting the portions of the OL and TS that are no longer applicable to a permanently defueled facility, while modifying the other sections to correspond to the permanently defueled condition.

The deletion and modification of provisions of the administrative controls do not directly affect the design of structures, systems, and components (SSCs) necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the fuel pool. The changes to the administrative controls are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shutdown and defueled condition of the reactor. The deletion of the Mitigation Strategy License Condition is also administrative in nature as the sections of the Order requiring implementation of the condition have been rescinded and the controlling regulation in which the mitigation strategies have been codified, 10 CFR 50.54(hh), specifies that these requirements are not applicable in the permanently defueled condition.

In a permanently defueled condition, the only credible accident is the fuel handling accident.

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

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

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

Response: No.

The proposed changes have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The administrative removal of an OL condition [* * *] or modifications of the TS that are related only to administration of facility cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled and VY will no longer [be] authorized to operate the reactor.

The proposed deletion of requirements of the VY OL and TS do not affect systems credited in the accident analysis for the fuel handling accident at VY. The proposed OL and TS will continue to require proper control and monitoring of safety significant parameters and activities.

The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding and spent fuel cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.Start Printed Page 9495

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.

Because the 10 CFR Part 50 license for VY will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel once the certifications required by 10 CFR 50.82(a)(1) are submitted, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The only remaining credible accident is a fuel handling accident (FHA). The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact the FHA.

The proposed changes are limited to those portions of the OL and TS that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the VY OL and TS are not credited in the existing accident analysis for the remaining applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated DBAs involving the reactor are no longer possible because the reactor will be permanently shutdown and defueled and VY will no longer be authorized to operate the reactor.

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: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, New York 10601.

NRC Branch Chief: Benjamin G. Beasley.

Indiana Michigan Power Company (IandM), Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

Date of amendment request: October 8, 2013.

Description of amendment request: The proposed amendment would increase the normal reactor coolant system (RCS) temperature and pressure at the Donald C. Cook Nuclear Plant, Unit 1, consistent with the previously licensed conditions. The proposed amendment would modify the Unit 1 technical specifications and license basis associated with this change.

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

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

Response: No.

  • SR 3.4.14.1 RCS [Pressure Isolation Valve (PIV)] Leakage—Surveillance Requirements

The proposed change to the RCS PIV RCS pressure range does not significantly increase the probability or consequences of an accident previously evaluated in the [Updated Final Safety Analysis Report (UFSAR)]. The analytical and evaluation efforts performed for the [Normal Operating Pressure/Normal Operating Temperature (NOP/NOT)] conditions were shown to be acceptable. The systems and components (including interface systems and control systems) will function as designed and all performance requirements for these systems remain acceptable. There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. No significant increase in the consequences has been identified. The NOP/NOT conditions do not introduce the possibility of a change in the frequency of an accident because the parameter changes are not an initiator of any accident previously considered and no new failure modes have been introduced.

Therefore, neither the probability nor the consequences of an accident previously evaluated has been significantly increased.

  • SR 3.5.5.1 Seal Injection Flow—Surveillance Requirements

The proposed change to the pressurizer pressure range and the elimination of the low pressure operation does not significantly increase the probability or consequences of an accident previously evaluated in the UFSAR. The analytical and evaluation efforts performed for the NOP/NOT conditions were shown to be acceptable. The systems and components (including interface systems and control systems) will function as designed and all performance requirements for these systems remain acceptable. There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. No significant increase in the consequences has been identified. The NOP/NOT conditions do not introduce the possibility of a change in the frequency of an accident because the parameter changes are not an initiator of any accident previously considered and no new failure modes have been introduced.

Therefore, neither the probability nor the consequences of an accident previously evaluated has been significantly increased.

  • SR 3.6.10.1 Containment Air Recirculation/Hydrogen Skimmer (CEQ) System—Surveillance Requirements

The proposed change to the containment air recirculation fan delay/start times does not significantly increase the probability or consequences of an accident previously evaluated in the UFSAR. The analytical and evaluation efforts performed for the NOP/NOT conditions were shown to be acceptable. The systems and components (including interface systems and control systems) will function as designed and all performance requirements for these systems remain acceptable. There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. No significant increase in the consequences has been identified. The NOP/NOT conditions do not introduce the possibility of a change in the frequency of an accident because the parameter changes are not an initiator of any accident previously considered and no new failure modes have been introduced.

Therefore, neither the probability nor the consequences of an accident previously evaluated has been significantly increased.

  • UFSAR Section 6.3.2, Containment Spray Systems [CTSs], System Design

The proposed revision to UFSAR Section 6.3.2 specifically recognizes use of the CTS pump time delay relay in mitigating the consequences of postulated accidents. Previously, the setting of this relay was established to support proper [emergency diesel generator] bus loading and it was accounted for as an input to accident analyses. Use of the time delay relay setting to mitigate the consequences of an accident does not significantly increase the probability or consequences of an accident previously evaluated in the UFSAR. The analytical and evaluation efforts performed for the NOP/NOT conditions were shown to be acceptable. The systems and components (including interface systems and control systems) will function as designed and all performance requirements for these systems remain acceptable. There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. No significant increase in the consequences has been identified. The NOP/NOT conditions do not introduce the possibility of a change in the frequency of an accident because the parameter changes are not an initiator of any accident previously considered and no new failure modes have been introduced.

Therefore, neither the probability nor the consequences of an accident previously evaluated has been significantly increased.

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

Response: No.

  • SR 3.4.14.1 RCS PIV Leakage—Surveillance Requirements

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated in the UFSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. This proposed change has no adverse effects on any safety related system and does not challenge the performance or integrity of any safety related system. The specified RCS pressure functions support meeting the accident analyses criteria.

Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.Start Printed Page 9496

  • SR 3.5.5.1 Seal Injection Flow—Surveillance Requirements

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated in the UFSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. This proposed change has no adverse effects on any safety related system and does not challenge the performance or integrity of any safety related system. The specified pressurizer pressure range supports meeting all of the accident analyses criteria.

Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.

  • SR 3.6.10.1 Containment Air Recirculation/Hydrogen Skimmer (CEQ) System—Surveillance Requirements

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated in the UFSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. This proposed change has no adverse effects on any safety related system and does not challenge the performance or integrity of any safety related system. The delay/start time functions support meeting all of the accident analyses criteria.

Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.

  • UFSAR Section 6.3.2, Containment Spray Systems, System Design

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated in the UFSAR because this change simply recognizes potential use of the existing CTS pump time delay relay setting to mitigate the consequences of an accident. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed change. This proposed change has no adverse effects on any safety related system and does not challenge the performance or integrity of any safety related system. The delay/start time functions support meeting all of the accident analyses criteria.

Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.

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

Response: No.

  • SR 3.4.14.1 RCS PIV Leakage—Surveillance Requirements

The proposed change does not involve a significant reduction in a margin of safety. Analyses and evaluations supporting the Return to NOP/NOT Program conditions demonstrate that all acceptance criteria continue to be met. There are no changes to the design, material, and construction standards that are applicable to any System, Structure, or Component (SSC). There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. Also, there is no change to a Design Basis Limit for Fission Product Barriers (DBLFPB).

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

  • SR 3.5.5.1 Seal Injection Flow—Surveillance Requirements

The proposed change does not involve a significant reduction in a margin of safety. Analyses and evaluations supporting the Return to NOP/NOT Program demonstrate that all acceptance criteria continue to be met. There are no changes to the design, material, and construction standards that are applicable to any SSC. There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. Also, there is no change to a DBLFPB.

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

  • SR 3.6.10.1 Containment Air Recirculation/Hydrogen Skimmer (CEQ) System—Surveillance Requirements

The proposed change does not involve a significant reduction in a margin of safety. Analyses and evaluations supporting the Return to NOP/NOT Program conditions demonstrate that all acceptance criteria continue to be met. There are no changes to the design, material, and construction standards that are applicable to the CEQ System. There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. Also, there is no change to a DBLFPB.

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

  • UFSAR Section 6.3.2, Containment Spray Systems, System Design

The proposed change does not involve a significant reduction in a margin of safety. There are no changes to the design, material, and construction standards that are applicable to the Containment Spray System. There are no physical changes being made to the fuel cladding, the RCS pressure boundary, or the containment. Also, there is no change to a DBLFPB.

Therefore, the proposed changes do 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.

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: November 6, 2013.

Description of amendment request: The proposed amendment would revise Technical Specification 3.6.13, Divider Barrier Integrity, concerning the divider barrier seal inspection requirements for the Donald C. Cook Nuclear Plant, Units 1 and 2.

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

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

Response: No.

The proposed changes do not involve changes to the installed structures, systems or components of the facility. The affected component (divider barrier seal) is not an accident initiator and therefore, this change does not involve a significant increase in the probability of an accident. The proposed change is considered adequate to ensure continued operability of the divider barrier. Since the divider barrier will continue to be available to perform its accident mitigation function, the consequences of accidents previously evaluated are not increased.

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

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

Response: No.

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

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

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

Response: No.

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

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

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.Start Printed Page 9497

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

Date of amendment request: December 20, 2013.

Description of amendment request: The proposed amendments would revise the Prairie Island Nuclear Generating Plant, Units 1 and 2, Emergency Plan to increase the staff augmentation times for certain Emergency Response Organization functions from 30 minutes and 60 minutes to 90 minutes.

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 increase in staff augmentation times has no effect on normal plant operation or on any accident initiator or precursors and does not impact the function of plant structures, systems, or components (SSCs). 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. The ability of the emergency response organization to respond adequately to radiological emergencies has been demonstrated as acceptable through a staffing analysis as required by 10 CFR Part 50, Appendix E.IV.A.9.

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

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

Response: No.

The proposed 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 increases the staff augmentation response times in the Emergency Plan, which are demonstrated as acceptable through a staffing analysis as required by 10 CFR Part 50, Appendix E.IV.A.9. 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.

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

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

Response: No.

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. A staffing analysis and a functional analysis were performed for the proposed change on the timeliness of performing major tasks for the functional areas of Emergency Plan. The analysis concluded that an increase in staff augmentation times, with the addition of two on-shift positions, would not significantly affect the ability to perform the required Emergency Plan tasks. Therefore, the proposed change is determined to not adversely affect the ability to meet 10 CFR 50.54(q)(2), the requirements of 10 CFR Part 50, Appendix E, and the emergency planning standards as described in 10 CFR 50.47 (b).

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

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

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

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: November 26, 2013.

Description of amendment request: The proposed change would amend Combined License Nos. NPF-93 and NPF-94 for VCSNS Units 2 and 3 by departing from approved AP1000 Design Control Document (DCD) Tier 2 information as incorporated into the Updated Final Safety Analysis Report (UFSAR) to allow use of a new methodology to determine the effective thermal conductivity resulting from oxidation of the inorganic zinc (IOZ) used in the containment vessel coating system.

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

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

Response: No.

Implementation of a methodology which specifies an effective thermal conductivity and oxidation progression for the inorganic zinc coating of the containment vessel is used to eliminate non-mechanistic modeling of inorganic zinc thermal conductivity in the containment integrity analyses to show that the value for inorganic zinc thermal conductivity used in the containment integrity analyses is conservative, but is not used to change any of the parameters used in those analyses. There is no change to any accident initiator or condition of the containment that would affect the probability of any accident. The containment peak pressure analysis as reported in the UFSAR is not affected; therefore, the previously reported consequences are not affected.

Therefore, there is no 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 to implement a methodology which specifies an effective thermal conductivity and oxidation progression and effects for the inorganic zinc coating of the containment vessel is used to eliminate non-mechanistic modeling of inorganic zinc thermal conductivity in the containment integrity analyses to show that the value for inorganic zinc thermal conductivity used in the containment integrity analyses is conservative, but is not used to change any of the parameters used in the containment peak pressure analysis. The change in methodology does not change the condition of containment; therefore, no new accident initiator is created. The containment peak pressure analysis as currently evaluated is not affected, and the consequences previously reported are not changed. The new methodology does not change the containment; therefore, no new fault or sequence of events that could lead to containment failure or release of radioactive material is created.

Therefore, this activity 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.Start Printed Page 9498

The proposed implementation of a methodology which specifies an effective thermal conductivity and oxidation progression and effects for the inorganic zinc coating of the containment vessel is used to eliminate non-mechanistic modeling of inorganic zinc thermal conductivity in the containment integrity analyses to show that the value for inorganic zinc thermal conductivity used in the containment integrity analyses is conservative, but is not used to change any of the parameters used in the containment peak pressure analysis. The change in methodology does not change the condition of the containment and the integrity of the containment vessel is not affected. The containment peak pressure analysis as currently evaluated is not affected, and the consequences previously reported are not changed. No safety analysis or design basis acceptance limit/criterion is changed by the proposed change, thus no margin of safety is reduced.

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

NRC Branch Chief: Lawrence Burkhart.

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

Date of amendment request: December 17, 2013.

Description of amendment request: The proposed amendment would revise the VCSNS Units 2 and 3 Emergency Plan to facilitate compliance with the Final Rule for Emergency Planning and Preparedness published on November 23, 2011. These proposed changes include the addition of text that (1) clarifies the distance of the Emergency Operations Facility from the site, (2) updates the content of exercise scenarios to be performed at least once each exercise cycle, and (3) requires the Evacuation Time Estimate to be updated annually between decennial censuses. This amendment request also proposes a new license condition to ensure the completion of a staffing analysis of on-shift personnel responsibilities no later than 180 days before fuel load.

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 VCSNS Units 2 and 3 Emergency Plan provides assurance that the requirements of emergency preparedness regulations are met. The changes do not affect the design, construction, or operation of the nuclear plant, so there is no change to the probability or consequences of an accident previously evaluated.

Adding a license condition related to an emergency preparedness staffing analysis and changing the VCSNS Units 2 and 3 Emergency Plan does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses as the purpose of the plan is to implement emergency preparedness regulations. No safety-related structure, system, component (SSC) or function is adversely affected. The change does not involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. Because the changes do not involve any SSC or function used to mitigate an accident, the consequences of the accidents evaluated in the UFSAR are not affected.

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

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

Response: No.

The VCSNS Units 2 and 3 Emergency Plan provides assurance that the requirements of emergency preparedness regulations are met. The changes do not affect the design, construction, or operation of the nuclear plant, so there is no new or different kind of accident from any accident previously evaluated. The changes do not affect safety-related equipment, nor do they affect equipment which, if it failed, could initiate an accident or a failure of a fission product barrier. In addition, the changes do not result in a new failure mode, malfunction, or sequence of events that could affect safety or safety-related equipment.

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

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

Response: No.

The VCSNS Units 2 and 3 Emergency Plan provides assurance that the requirements of emergency preparedness regulations are met. The changes do not affect the assessments or the plant itself. The changes do not affect safety-related equipment or equipment whose failure could initiate an accident, nor does it adversely interface with safety-related equipment or fission product barriers. No safety analysis or design basis acceptance limit or criterion is challenged or exceeded by the requested change.

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

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

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

NRC Branch Chief: Lawrence Burkhart.

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 Start Printed Page 9499Assessment as indicated. All of these items are available for public inspection at the NRC's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through the Agencywide Documents Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr.resource@nrc.gov.

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland

Date of application for amendments: January 28, 2013, as supplemented by letter dated April 1, 2013.

Brief description of amendments: The amendments revised Technical Specification (TS) 1.3, “Completion Times” Example 1.3-3, TS 3.6.6, “Containment Spray and Cooling Systems,” TS 3.7.3, “Auxiliary Feedwater (AFW) System,” TS 3.8.1, “AC [Alternating Current] Sources-Operating,” and TS 3.8.9, “Distribution Systems-Operating” by eliminating the second completion time in accordance with TS Task Force (TSTF)-439-A, Revision 2, “Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO [limiting condition for operation].”

Date of issuance: January 29, 2014.

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

Amendment Nos.: 304 and 282.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: May 28, 2013 (78 FR 31981).

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated January 29, 2014.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit 2 (MPS2), New London County, Connecticut

Date of amendment request: March 21, 2013.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.1.3.7—Control Rod Drive Mechanisms to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal for MPS2.

Date of issuance: January 30, 2014.

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

Amendment No.: 317.

Renewed Facility Operating License No. DPR-65: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: June 11, 2013 (78 FR 35061).

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

No significant hazards consideration comments received: No.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of application for amendment: January 11, 2013.

Brief description of amendment: The amendment revises the Fermi 2 Technical Specifications (TSs) to risk-inform requirements regarding selected Required Action end states. Additionally, it would modify the TSs Required Actions with a Note prohibiting the use of limiting condition for operation 3.0.4.a when entering the preferred end state (Mode 3) on startup. The changes are consistent with the NRC's Technical Specification Task Force traveler TSTF-423, Revision 1, “Technical Specifications End States, NEDC-32988-A,” dated December 22, 2009 (ADAMS Accession No. ML093570241).

Date of issuance: January 17, 2014.

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

Amendment No.: 194.

Facility Operating License No. NPF-43: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: April 16, 2013 (78 FR 22565).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 31, 2012, as supplemented by letters dated July 31, August 22, October 5, and November 12, 2012, and January 7, April 11, May 9, and August 6, 2013.

Brief description of amendment: The amendment allows the licensee to expand the operating domain by the implementation of Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Power Range Neutron Monitoring/Maximum Extended Load Line Limit Analysis (ARTS/PRNM/MELLLA). The Neutron Monitoring System will be modified by replacing the Average Power Range Monitor (APRM) subsystem with the Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) System. The modification of the PRNM system replaces analog technology with digital technology to improve the management and maintenance of the system. The licensee will expand the operating domain to Maximum Extended Load Line Limit Analysis (MELLLA) and make changes to certain allowable values and limits and to the Technical Specifications (TSs). The changes to the TSs include the adoption of Technical Specifications Task Force (TSTF) Change Traveler TSTF-493, “Clarify Application of Setpoint Methodology for LSSS [Limiting Safety System Setting] Functions,” Option A surveillance notes. Furthermore, the amendment allows a change in the licensing basis to support Anticipated Transient without Scram accident mitigation with one Standby Liquid Control pump instead of two.

Date of Issuance: January 31, 2014.

Effective Date: The license amendment is effective as of its date of issuance and shall be implemented within 60 days thereafter. The Technical Specification revisions will be applicable following completion of the refueling outage (R22) scheduled to begin May 8, 2015.

Amendment No.: 226.

Renewed Facility Operating License No. NPF-21: The amendment revised the Facility Operating License and Technical Specifications.

Date of Initial Notice in Federal Register: September 11, 2012 (77 FR 55867). The supplemental letters dated July 31, August 22, October 5, and November 12, 2012, and January 7, April 11, May 9, and August 6, 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 amendment is contained in a Start Printed Page 9500Safety Evaluation dated January 31, 2014.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 18, 2012, as supplemented on March 12, 2013, July 17, 2013, and November 15, 2013.

Brief description of amendment: The amendment revises the MNGP Renewed Facility Operating License and Technical Specification (TS) 3.8.3, “Diesel Fuel Oil, Lube Oil, and Starting Air,” by removing the current stored diesel fuel oil, and lube oil numerical volume requirements from the TSs and replacing them with duration-based numerical requirements consistent with TSTF-501, Revision 1.

Date of issuance: January 28, 2014.

Effective date: This license amendment is effective as of the date of issuance and shall be implemented within 60 days from date of issuance.

Amendment No.: 178.

Renewed Facility Operating License No. DPR-22: Amendment revises the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: December 11, 2012 (77 FR 73689). The licensee's supplements dated March 12, 2013, July 17, 2013, and November 15, 2013, did not change the scope of the original amendment request, did not change the NRC staff's initial proposed finding of no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 2014.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 13, 2012, as supplemented by letters dated June 21, 2013, and July 23, 2013.

Brief description of amendments: The amendments made changes to the Prairie Island Nuclear Generating Plant Emergency Plan emergency action level initiating conditions for the classification of liquid effluent releases and for the determination of fuel clad barrier loss.

Date of issuance: January 25, 2014.

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

Amendment Nos.: Unit 1—210; Unit 2—198.

Renewed Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Prairie Island Nuclear Generating Plant Emergency Plan.

Date of initial notice in Federal Register: March 4, 2013 (78 FR 14134). The supplemental letters dated June 21, 2013, and July 23, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 25, 2014.

No significant hazards consideration comments received: No.

Northern States Power Company—Minnesota, Docket Nos. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: December 21, 2012, as supplemented by letter dated May 16, 2013.

Brief description of amendments: The amendment made changes to the Monticello Nuclear Generating Plant Emergency Plan by revising the Emergency Action Level (EAL) setpoint for the Turbine Building Normal Waste Sump (TBNWS) Monitor. The change to the EAL restores indication of an Alert classification of a liquid effluent release via the TBNWS pathway to within the indication range of the applicable instrumentation.

Date of issuance: January 28, 2014.

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

Amendment No.: 177.

Renewed Facility Operating License No. DPR-22: Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: March 4, 2013 (78 FR 14133). The supplemental letter dated May 16, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 2014.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 27, 2012, as supplemented by letter dated June 27, 2013.

Brief description of amendment: The amendment revised the Fort Calhoun Station, Unit 1 (FCS) Technical Specification (TS) Limiting Condition for Operation 2.16, “River Level,” and TS Surveillance Requirement 3.2, “Equipment and Sampling Tests,” and a related change to the FCS Radiological Emergency Response Plan to revise two emergency action levels related to high water level in the Missouri River.

Date of issuance: January 28, 2014.

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

Amendment No.: 274.

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

Date of initial notice in Federal Register: December 26, 2012 (77 FR 76082). The supplemental letter dated June 27, 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 amendment is contained in a safety evaluation dated January 28, 2014.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, Fairfield County, South Carolina

Date of application for amendment: April 3, 2013.

Brief description of amendment: This amendment allows for the extension of the 130-month frequency of the VCSNS containment integrated leak rate test (ILRT) or Type A test, that is required by TS 6.8.4(g) to 15 years on a permanent basis.

Date of issuance: February 5, 2014.

Effective date: This license amendment is effective as of the date of its issuance.

Amendment No.: 194.Start Printed Page 9501

Renewed Facility Operating License No. NPF-12: Amendment revises the License.

Date of initial notice in Federal Register: June 25, 2013 (78 FR 38084).

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

No significant hazards consideration comments received: No.

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

Date of amendment request: July 15, 2013, as supplemented by a letter dated November 15, 2013.

Brief description of amendment: The proposed amendment modified design details related to the construction of Module CA03 which forms the west wall of the in-containment refueling water storage tank. The changes sought to clarify the materials used in fabrication of the module, as well as the design details related to the horizontal stiffeners used to support the in-containment refueling water storage tank, and module legs used to anchor the module in place.

Date of issuance: January 28, 2014.

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

Amendment No.: Unit 3-17, and Unit 4-17.

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

Date of initial notice in Federal Register: September 3, 2013 (78 FR 54288).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 2014.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 10th day of February 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. 2014-03494 Filed 2-18-14; 8:45 am]

BILLING CODE 7590-01-P