2013-20154. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

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 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 July 25, 2013 to August 7, 2013. The last biweekly notice was published on August 6, 2013 (78 FR 47785).

ADDRESSES:

You may submit comment 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-2013-0191. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail].
• Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: 3WFN, 06A44M, 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.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2013-0191 when contacting the NRC about the availability of information regarding this document. You may access publicly-available information related to this action by the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0191.
• 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 [e-mail]. Documents may be viewed in ADAMS by performing a search on the document date and docket number.
• 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-2013-0191 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

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

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

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

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

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 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.

All documents filed in the NRC's 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 [e-mail], 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 agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

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

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC 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 to [e-mail], 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 following 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's 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 [e-mail].

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South Carolina

Date of amendment request: June 7, 2013.

Description of amendment request: The proposed change would delete the current HBRSEP Surveillance Requirements (SRs) 3.1.7.1, 3.1.7.2, and 3.1.7.3 of Technical Specification 3.1.7, “Rod Position Indication,” and renumber current SR 3.1.7.4 as SR 3.1.7.1. This change deletes a redundant SR and eliminates a minimum of eight reactivity manipulations per year.

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 initiating conditions and assumptions for dose consequences of accidents described in the Updated Final Safety Analyses Report remain as previously analyzed. The proposed change does not introduce a new accident initiator nor does it introduce changes to any existing accident initiators described in the Updated Final Safety Analyses Report. The proposed change eliminates requirements to periodically demonstrate agreement of individual rod position with average rod position and group demand step counter position during control rod movement while maintaining less frequent requirements for control rod movement associated with verification of control rod freedom of movement (SR 3.1.4.2) and confirmation that the two rod position indication systems are within alignment limits (SR 3.1.4.1). Control rod movement is a potential accident initiator and less frequent surveillances involving less control rod movement will not increase the probability or consequences of an accident.

The proposed change also eliminates surveillance requirements which are redundant to the requirements of SR 3.1.4.1 and modifies SR 3.1.7.4 to renumber it as SR 3.1.7.1. The elimination of redundant surveillance requirements does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Current SR 3.1.7.4 and proposed SR 3.1.7.1 involve the maintenance and configuration of instrumentation used to indicate rod position. The proposed change renumbers SR 3.1.7.4 as SR 3.1.7.1 and maintains the requirement to perform a Channel Calibration on an 18 Month-Frequency which does not change the means and manner of control of control rod movement and therefore does not involve a significant increase in the probability of consequences of an accident previously evaluated.

Based on the above, 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 previously evaluated?

Response: No.

The proposed change will not introduce any new failure modes to the required protection functions. The proposed change modifies surveillance requirements associated with operation and function of instrumentation indicating rod position that is part of the control rod control system (demand step counter position) and individual analog rod position indication instrumentation. The proposed change does not alter the manner in which the respective rod position indications function or the control system controls control rod movement such that the modified surveillance requirements of TS 3.1.7 cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

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 the margin of safety?

Response: No.

The proposed amendment does not involve revisions to any safety analysis limits or safety system settings that will adversely impact plant safety. The proposed amendment does not alter the functional capabilities assumed in a safety analysis for any system, structure, or component important to the mitigation and control of design bases accident conditions within the facility. Nor does this amendment revise any parameters or operating restrictions that are assumptions of a design basis accident. In addition, the proposed amendment does not affect the ability of safety systems to ensure that the facility can be placed and maintained in a shutdown condition for extended periods of time.

The Technical Specifications continue to assure that the applicable operating parameters and systems are maintained within the design requirements and safety analysis assumptions. Therefore, the proposed changes which eliminate surveillance requirements that are either redundant or inconsistent with industry standards for the partial movement of control rods and rod position indication system surveillance and add a new requirement that the rod position indication systems agree within a prescribed value will not result in a significant reduction in the margin of safety as defined in the Updated Final Safety Analyses Report or 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.

Attorney for licensee: David T. Conley, Manager—Senior Counsel—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Acting Branch Chief: Douglas A. Broaddus.

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: April 17, 2013.

Description of amendment request: The proposed amendment would modify the Fermi 2 technical specification (TS) related to control room envelope habitability in accordance with NRC-approved Technical Specifications Task Force (TSTF) change traveler TSTF-448, “Control Room Habitability,” Revision 3. The proposed amendment is consistent with the Consolidated Line Item Improvement Process that adopts changes to TS Section 3.7.3, “Control Room Emergency Filtration (CREF) System,” and adds TS Section 5.5.14, “Control Room Envelope Habitability Program.”

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration, is presented below. The licensee incorporated, by reference, the proposed no significant hazards consideration published in the Federal Register on January 9, 2007 (72 FR 2032).

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

Response: No.

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident 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 impact the accident analysis. The proposed change does not alter the required mitigation capability of the CRE emergency ventilation system, or its functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice.

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

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

Response: No.

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

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

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

Attorney for licensee: Bruce R. Masters, DTE Energy, General Counsel—Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: April 16, 2013.

Description of amendment request: The proposed amendment would revise the Renewed Facility Operating License by deleting a license condition associated with license renewal.

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 would modify the KPS renewed facility operating license by deleting a license condition that pertains to plant operation during the period of extended operation. KPS is permanently ceasing operation and will permanently defuel the reactor vessel prior to the start of the period of extended operation. Therefore, the probability of occurrence of previously evaluated accidents is not affected, since the original license did not contain this license condition. The license condition being deleted pertains to operation beyond the term of the original license. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor.

Since KPS is permanently ceasing operation, the generation of fission products will cease and the remaining source term will decay. This significantly reduces the consequences of the remaining applicable postulated accident. Therefore, the proposed amendment does not involve a significant increase in the consequences of a previously evaluated accident.

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

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 activities and programs that were the subject of this license condition were intended to ensure that systems, structures, and components (SSCs) continue to respond properly in the event of a previously analyzed accident during the period of extended operation of the renewed facility operating license. However, the reactor will not operate during the period of extended operation.

The proposed amendment does not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed amendment. Similarly, the proposed amendment would not physically change any SSCs involved in the mitigation of any postulated accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed amendment does not create the possibility of a new failure mode associated with any equipment or personnel failures.

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

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

Response: No.

Because the 10 CFR part 50 license for KPS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The remaining credible accident (90 days after shutdown) is a fuel handling accident (FHA) in the auxiliary building. The proposed amendment does not affect the inputs or assumptions of any of the design basis analyses that impact a FHA in the auxiliary building and the current design limits continue to be met for the accident of concern.

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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: May 29, 2013.

Description of amendment request: The proposed amendment would revise the operating license and revise the associated technical specifications (TSs) to the permanently defueled technical specifications (PDTSs) consistent with the permanent cessation of reactor operation.

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.

KPS has permanently ceased operation. The proposed amendment would modify the KPS renewed facility operating license and TS by deleting the portions of the license and TS that are no longer applicable to a permanently defueled facility, while modifying the remaining portions to correspond to the permanently shutdown condition. This change is consistent with the Standard TS and with the criteria set forth in 10 CFR 50.36 for the contents of TS.

Section 14 of the KPS Updated Safety Analysis Report (USAR) described the design basis accident (DBA) and transient scenarios applicable to KPS during power operations. With the reactor in a permanently defueled condition, the spent fuel pool and its systems have been isolated and are dedicated only to spent fuel storage. In this condition the spectrum of credible accidents is much smaller than for an operational plant. As a result of the certifications submitted by DEK in accordance with 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactor or to place or retain fuel in the reactor in accordance with 10 CFR 50.82(a)(2), most of the accident scenarios postulated in the USAR are no longer possible.

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

1. The integrity of the reactor coolant boundary;

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

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.43(a)(1) or 100.11.

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after the termination of reactor operations at KPS and the permanent removal of the fuel from the reactor vessel (following 90 days of decay time after shutdown) and purging of the contents of the waste gas decay tanks and liquid waste tanks, none of the SSCs at KPS are required to be relied on for accident mitigation. Therefore, none of the SSCs at KPS meet the definition of a safety-related SSC stated in 10 CFR 50.2 (with the exception of the passive spent fuel pool structure).

The deletion of TS definitions and rules of usage and application, that are currently not applicable in a defueled condition, has no impact on facility SSCs or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shutdown and defueled status of KPS has no impact on the remaining DBA (the fuel handling accident in the auxiliary building). The removal of limiting conditions for operation (LCOs) or surveillance requirements (SRs) that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated since these DBAs are no longer applicable in the defueled mode. The safety functions involving core reactivity control, reactor heat removal, reactor coolant system inventory control, and containment integrity are no longer applicable at KPS as a permanently defueled plant. The analyzed accidents involving damage to the reactor coolant system, main steam lines, reactor core, and the subsequent release of radioactive material are no longer possible at KPS.

Since KPS has permanently ceased operation, the future generation of fission products has ceased and the remaining source term will decay. The radioactive decay of the irradiated fuel since shutdown of the reactor will have reduced the consequences of the fuel handling accident to levels well below those previously analyzed. The relevant parameter (water level) associated with the fuel pool provides an initial condition for the fuel handling accident analysis and is included in the permanently defueled TS.

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

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

The deletion and modification of provisions of the administrative controls do not directly affect the design of SSCs necessary for 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 probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition is the only operation currently 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. These changes are consistent with the standard TS. The removal of TS that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor is permanently shutdown and defueled and KPS is no longer authorized to operate the reactor.

The proposed deletion of requirements of the KPS TS do not affect systems credited in the accident analysis for the fuel handling accident in the auxiliary building at KPS. The proposed permanently defueled TS (PDTS) continue to require proper control and monitoring of safety significant parameters and activities.

The proposed restriction on the fuel pool level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA. The spent fuel pool water level, spent fuel pool boron concentration, and spent fuel pool storage LCOs are retained to preserve the current requirements for safe storage of irradiated fuel.

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

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

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

Response: No

Because the 10 CFR Part 50 license for KPS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation 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 a FHA.

The proposed changes are limited to those portions of TS and license that are not related to the safe storage of irradiated fuel. The requirements for SSCs that have been deleted from the KPS 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 is permanently shutdown and defueled and KPS is no longer authorized to operate the reactor.

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

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

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: May 3, 2013.

Description of amendment request: The proposed amendment would revise the Millstone Power Station, Unit 2 (MPS2) Technical Specification (TS) 3/4.7.11, “Ultimate Heat Sink”, to increase the current ultimate heat sink water temperature limit from 75 °F to 80 °F and change the TS Action to state, “With the ultimate heat sink water temperature greater than 80 °F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.”

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

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

Response: No.

Previously evaluated accident consequences are not impacted because credited mitigating equipment continues to perform its design function. The proposed change does not significantly impact the probability of an accident previously evaluated because those SSCs that can initiate an accident are not significantly impacted.

Based on the above, DNC concludes that the proposed increased temperature limits do not involve a significant increase in the probability or consequences of an accident or transient previously evaluated in the safety analysis report.

Criterion 2

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

Response: No.

A new or different accident from any accident previously evaluated is not created because previously credited SSCs, are not impacted, there is no new reliance upon equipment not previously credited, there is no new equipment installed (except for monitoring equipment), there is no impact upon the existing failure modes and effects analysis, and conformance to the single failure criterion is maintained. The increased limits do not introduce any new mode of plant operation and will not result in a change to the design function or the operation of any SSC that is used for mitigating accidents.

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

Criterion 3

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

Response: No.

This change does not involve a significant reduction in margin of safety because the containment analysis acceptance criteria continue to be met when operating with the proposed increased UHS temperature limit. Containment integrity will not be challenged and will continue to meet its design basis acceptance criteria following a large break LOCA or MSLB. The proposed change has no impact upon fuel cladding or RCS fission product barrier margin because credited SSCs continue to perform their design functions with an 80 °F UHS temperature.

Based on the above, DNC concludes that the proposed changes do not involve a significant reduction in the margin of safety.

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

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

NRC Acting Branch Chief: Robert H. Beall.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit 3, New London County, Connecticut

Date of amendment request: May 3, 2013.

Description of amendment request: The proposed amendment would revise the Millstone Power Station, Unit 3 (MPS3) Technical Specification (TS) 3/4.7.5, “Ultimate Heat Sink”, to increase the current ultimate heat sink water temperature limit from 75 °F to 80 °F and change the TS Action to state, “With the ultimate heat sink water temperature greater than 80 °F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.”

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

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

Response: No.

Previously evaluated accident consequences are not impacted because credited mitigating equipment continues to perform its design function. The proposed change does not significantly impact the probability of an accident previously evaluated because those SSCs that can initiate an accident are not significantly impacted.

Based on the above, DNC concludes that the proposed increased temperature limits do not involve a significant increase in the probability or consequences of an accident or transient previously evaluated in the safety analysis report.

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

Response: No.

A new or different accident from any accident previously evaluated is not created because previously credited SSCs, are not impacted; there is no new reliance upon equipment not previously credited; there is no new equipment installed (except for monitoring equipment); there is no impact upon the existing failure modes and effects analysis; and conformance to the single failure criterion is maintained.

The increased limits do not introduce any new mode of plant operation and will not result in a change to the design function or the operation of any SSC that is used for mitigating accidents.

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

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

Response: No.

This change doesn't involve a significant reduction in margin of safety because containment structure fission product barrier design margin is unaffected because peak pressure/temperature occurs early in the accident before UHS temperature can influence the containment response. The proposed change has no impact upon fuel cladding or RCS fission product barrier margin because credited SSCs continue to perform their design functions with an 80 °F UHS temperature.

Based on the above, DNC concludes that the proposed changes do not involve a significant reduction in the margin of safety.

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

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

NRC Acting Branch Chief: Robert H. Beall.

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: June 13, 2013.

Description of amendment request: The amendment will adopt Technical Specification Task Force (TSTF)-423, Revision 1, “Technical Specifications End States.” Specifically, the proposed amendment would modify Technical Specifications (TSs) to risk-informed requirements regarding selected Required Action end states. The proposed changes are consistent with NRC-approved TSTF-423, Revision 1, with some deviations noted.

The NRC issued a “Notice of Availability of the Proposed Models for Plant-Specific Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-423, Revision 1, `Technical Specifications End States, NEDC-32988-A,' for Boiling Water Reactor Plants Using the Consolidated Line Item Improvement Process,” published in the Federal Register on February 18, 2011 (76 FR 9614), which included the model no significant hazards consideration and safety evaluation for TSTF-423, Revision 1.

Basis for proposed no significant hazards consideration determination: An analysis of the no significant hazards consideration was presented in the TSTF-423. The licensee has affirmed the applicability of the model no significant hazards consideration determination, 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.

Required Actions are not an initiator of any accident previously evaluated. Therefore, the proposed changes do not affect the probability of any accident previously evaluated. NEDC-32988-A demonstrated that the proposed changes in the required end state do not significantly increase the consequences of any accidents previously evaluated.

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

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

Response: No.

The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis.

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

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

Response: No.

NEDC-32988-A demonstrated that the changed end states represent a condition of equal or lower risk than the original end states.

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

Based on the above, the TSTF-423 concludes that the proposed change 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.

Attorney for licensee: Joseph A. Aluise, Associate General Counsel—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.

NRC Branch Chief: Michael T. Markley.

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: July 16, 2013.

Description of amendment request: The amendment would adopt Technical Specifications Task Force (TSTF) change traveler TSTF-535, Revision 0, “Revise Shutdown Margin Definition to Address Advanced Fuel Designs.” The Shutdown Margin (SDM) (i.e., the amount of reactivity by which the reactor is subcritical) is calculated under the conservative conditions that the reactor is Xenon free, the most reactive control rod is outside the reactor core, and the moderator temperature produces the maximum reactivity. For standard fuel designs, maximum reactivity occurs at a moderator temperature of 68 degrees Fahrenheit (°F), which is reflected in the temperature specified in the Technical Specifications (TSs). New, advanced Boiling Water Reactor (BWR) fuel designs can have a higher reactivity at moderator shutdown temperatures above 68 °F. Therefore, the proposed amendment, consistent with TSTF-535, Revision 0, seeks to modify the TSs to require the SDM to be calculated at whatever temperature produces the maximum reactivity (i.e., temperatures at or above 68 °F).

The notice of availability of this TS improvement “Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF-535, Revision 0, `Revise Shutdown Margin Definition to Address Advanced Fuel Designs,' Using the Consolidated Line Item Improvement Process,” was published in Federal Register on February 26, 2013 (78 FR 13100), which included a model no significant hazards consideration (NSHC) determination and safety evaluation.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has affirmed the applicability of the model no significant hazards consideration determination included in TSTF-535, Revision 0, and provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change revises the definition of SDM. SDM is not an initiator to any accident previously evaluated. Accordingly, the proposed change to the definition of SDM has no effect on the probability of any accident previously evaluated. SDM is an assumption in the analysis of some previously evaluated accidents and inadequate SDM could lead to an increase in consequences for those accidents. However, the proposed change revises the SDM definition to ensure that the correct SDM is determined for all fuel types at all times during the fuel cycle. As a result, the proposed change does not adversely affect the consequences of any accident previously evaluated.

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

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

Response: No.

The proposed change revises the definition of SDM. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operations. The change does not alter assumptions made in the safety analysis regarding SDM.

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

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

Response: No.

The proposed change revises the definition of SDM. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change ensures that the SDM assumed in determining safety limits, limiting safety system settings or limiting conditions for operation is correct for all BWR fuel types at all times during the fuel cycle.

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: Joseph A. Aluise, Associate General Counsel—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: March 22, 2013.

Description of amendment request: The license amendment request proposes to revise the Technical Specifications (TS) to allow the use of Optimized ZIRLO^TM fuel rod cladding material. The proposed change would revise TS 5.3.1 to add Optimized ZIRLO^TM to the approved fuel rod cladding materials and TS 6.9.1.7 to add Westinghouse Electric Company LLC topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO^TM,” to the analytical methods used to determine the core operating limits.

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

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

Response: No.

The proposed change would allow the use of Optimized ZIRLO^TM clad nuclear fuel in the reactors. The NRC approved topical report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A “Optimized ZIRLO^TM,” prepared by Westinghouse Electric Company LLC (Westinghouse), addresses Optimized ZIRLO^TM and demonstrates that Optimized ZIRLO^TM has essentially the same properties as currently licensed ZIRLO.® The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO^TM fuel cladding will continue to meet all 10 CFR 50.46 acceptance criteria and, therefore, will not increase the consequences of an accident.

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.

Use of Optimized ZIRLO^TM clad fuel will not result in changes in the operation or configuration of the facility. Topical Report WCAP-12610-PA and CENPD-404-P-A demonstrated that the material properties of Optimized ZIRLO^TM are similar to those of standard ZIRLO.® Therefore, Optimized ZIRLO^TM fuel rod cladding will perform similarly to those fabricated from standard ZIRLO,® thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident.

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 the margin of safety?

Response: No.

The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the Optimized ZIRLO^TM are not significantly different from those of standard ZIRLO.® Optimized ZIRLO^TM is expected to perform similarly to standard ZIRLO® for all normal operating and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference in Optimized ZIRLO^TM material properties relative to standard ZIRLO^TM could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO properties demonstrates that the acceptance criteria of 10 CFR 50.46 has been satisfied.

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: James Petro, Managing Attorney—Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Acting Branch Chief: Douglas A. Broaddus.

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

Date of amendment request: May 28, 2013.

Description of amendment request: The proposed amendment will modify the Seabrook Technical Specifications (TSs). Specifically, the proposed amendment will modify the TS by relocating specific surveillance frequencies to a licensee-controlled program with implementation of Nuclear Energy Institute 04-10, “Risk-Informed Technical Specification Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies.” The changes are consistent with NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) change TSTF-425, “Relocate Surveillance Frequencies to Licensee Control—Risk Informed Technical Specifications Task Force (RITSTF) Initiative 5b,” Revision 3, (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of this TSTF improvement, and included a model no significant hazards consideration and safety evaluation.

Basis for proposed no significant hazards consideration determination: An analysis of the no significant hazards consideration was presented in the TSTF-425. The licensee has affirmed the applicability of the model no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to an any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

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

Response: No.

No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, NextEra will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

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: Mr. James Petro, Managing Attorney, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Acting Branch Chief: Veronica Rodriguez.

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

Date of amendment request: June 25, 2013.

Description of amendment request: The proposed amendment will revise the Seabrook Technical Specifications. Specifically, the proposed amendment will allow the use of Optimized ZIRLO^TM as fuel rod cladding.

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

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

Response: No.

The proposed change would allow the use of Optimized ZIRLO^TM clad nuclear fuel in the reactors. The NRC approved topical report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A “Optimized ZIRLO,^TM” prepared by Westinghouse Electric Company LLC (Westinghouse), addresses Optimized ZIRLO^TM and demonstrates that Optimized ZIRLO^TM has essentially the same properties as currently licensed ZIRLO.® The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO^TM fuel cladding will continue to meet all [Title 10 of the Code of Federal Regulations] 10 CFR 50.46 acceptance criteria and, therefore, will not increase the consequences of an accident.

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.

Use of Optimized ZIRLO^TM clad fuel will not result in changes in the operation or configuration of the facility. Topical Report WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material properties of Optimized ZIRLO^TM are similar to those of standard ZIRLO.® Therefore, Optimized ZIRLO^TM fuel rod cladding will perform similarly to those fabricated from standard ZIRLO,® thus precluding the possibility of the fuel cladding becoming an accident initiator and causing a new or different type of accident.

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 changes involve a significant reduction in a margin of safety?

Response: No.

The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the Optimized ZIRLO^TM are not significantly different from those of standard ZIRLO.® Optimized ZIRLO^TM is expected to perform similarly to standard ZIRLO® for all normal operating and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference in Optimized ZIRLO^TM material properties relative to standard [ZIRLO®], ZIRLO^TM could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO^TM properties will demonstrate that the acceptance criteria of 10 CFR 50.46 have been satisfied.

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

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

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

NRC Acting Branch Chief: Veronica Rodriguez.

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: February 2, 2013, as supplemented by letter dated June 25, 2013.

Description of amendment request: The proposed amendments would remove Technical Specification (TS) 3.5.3 “[Emergency Core Cooling Systems (ECCS)]—Shutdown” Limiting Condition for Operation (LCO) Note 1 to eliminate information to the plant operators that could cause non-conservative operation.

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

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

Response: No.

This license amendment request proposes to revise the Technical Specification for ECCS operability requirements in Mode 4 by removing the LCO Note which allows the RHR subsystem to be considered operable for ECCS when aligned for shutdown cooling. These changes will require one train of RHR to be aligned for ECCS operation throughout the mode and other specified conditions of applicability.

The proposed changes do not affect the ECCS and RHR subsystem design, the interfaces between the RHR subsystem and other plant systems' operating functions, or the reliability of the RHR subsystem. The proposed changes do not change or impact the initiators and assumptions of the analyzed accidents. Therefore, the ECCS and RHR subsystems will be capable of performing their accident mitigation functions, and the proposed removal of the LCO Note does not involve an increase in the probability of an accident.

The proposed removal of the LCO Note will require that one train of RHR is aligned for ECCS operation during the mode and other specified conditions of applicability which assures that one train of ECCS is operable to mitigate the consequences of a loss of coolant accident. Thus the proposed removal of the LCO Note does not involve a significant increase in the consequences of an accident.

Therefore, the proposed Technical Specification 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.

This license amendment request proposes to revise the Technical Specification for ECCS operability requirements in Mode 4 by removing the LCO Note which allows the RHR subsystem to be considered operable for ECCS when aligned for shutdown cooling. These changes will require one train of RHR to be aligned for ECCS operation throughout the mode and other specified conditions of applicability.

The proposed Technical Specification changes to remove the LCO Note involve changes to when system trains are operated, but they do not change any system functions or maintenance activities. The changes do not involve physical alteration of the plant, that is, no new or different type of equipment will be installed. The changes do not alter assumptions made in the safety analyses but ensure that one train of ECCS is operable to mitigate the consequences of a loss of coolant accident. These changes do not create new failure modes or mechanisms which are not identifiable during testing and no new accident precursors are generated.

Therefore, the proposed Technical Specification 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.

This license amendment request proposes to revise the Technical Specification for ECCS operability requirements in Mode 4 by removing the LCO Note which allows the RHR subsystem to be considered operable for ECCS when aligned for shutdown cooling. These changes will require one train of RHR to be aligned for ECCS operation throughout the mode and other specified conditions of applicability.

This license amendment proposes Technical Specification changes which assure that the ECCS—Shutdown TS LCO requirements are met if a Mode 4 LOCA were to occur. With these changes, other TS requirements for shutdown cooling in Mode 4 will continue to be met. Based on review of plant operating experience, there is no [discernible] change in cooldown rates when utilizing a single train of RHR for shutdown cooling. Thus, no margin of safety is reduced as part of this change.

Therefore, the proposed Technical Specification 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 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.

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: May 23, 2013.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plant, Units 1 and 2, to add a methodology to TS 5.6.5 “Core Operating Limits Report (COLR).”

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

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

Response: No.

This license amendment request proposes to revise the Technical Specifications to reference and allow use of WCAP-16045-P-A, “Qualification of the Two-Dimensional Transport Code PARAGON”, and WCAP-16045-P-A, Addendum 1-A, “Qualification of the NEXUS Nuclear Data Methodology”, for determining core operating limits.

The methodologies which this license amendment proposes for determination of core operating limits are improvements over the current methodologies in use at the Prairie Island Nuclear Generating Plant.

The NRC staff reviewed and approved these methodologies and concluded that these analysis codes are acceptable as a replacement for the current analysis code. Thus core operating limits determined using the proposed codes continue to assure that the reactor operates safely and, thus, the proposed changes do not involve an increase in the probability of an accident.

Operation of the reactor with core operating limits determined by use of the proposed analysis codes does not increase the reactor power level, does not increase the core fission product inventory, and does not change any transport assumptions. Therefore the proposed methodology and Technical Specification changes do not involve a significant increase in the consequences of an accident.

Therefore, the proposed methodology change and associated Technical Specification 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.

This license amendment request proposes to revise the Technical Specifications to reference and allow use of WCAP-16045-P-A, “Qualification of the Two-Dimensional Transport Code PARAGON”, and WCAP-16045-P-A, Addendum 1-A, “Qualification of the NEXUS Nuclear Data Methodology,” for determining core operating limits.

The proposed changes provide revised methodology for determining core operating limits, but they do not change any system functions or maintenance activities. The changes do not involve physical alteration of the plant, that is, no new or different type of equipment will be installed. The changes do not alter assumptions made in the safety analyses but ensure that the core will operate within safe limits. These changes do not create new failure modes or mechanisms which are not identifiable during testing, and no new accident precursors are generated.

Therefore, the proposed methodology change and associated Technical Specification 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.

This license amendment request proposes to revise the Technical Specifications to reference and allow use of WCAP-16045-P-A, “Qualification of the Two-Dimensional Transport Code PARAGON”, and WCAP-16045-P-A, Addendum 1-A, “Qualification of the NEXUS Nuclear Data Methodology,” for determining core operating limits.

This license amendment proposes revised methodology for determining core operating limits. The proposed methodology is an improvement that allows more accurate modeling of core performance. The NRC has reviewed and approved this methodology for use in lieu of the current methodology, thus, the margin of safety is not reduced due to this change.

Therefore, the proposed methodology change and associated Technical Specification 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 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.

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

Date of amendment request: April 12, 2013.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) 5.9.2. “Annual Radiological Environmental Operating Report,” to delete the reference to collocated dosimeters in relation to the NRC thermo luminescent dosimeters program. This change is consistent with NRC-approved Technical Specification Task Force (TSTF) change TSTF-348. In addition, it would correct a cross-reference error in TS 5.9.8, “Post Accident Monitoring System (PAMS) Report.”

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

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

Response: No.

The proposed changes do not require physical changes to plant systems, structures, or components. The proposed changes are administrative in nature and therefore, do not change the fundamental requirements of the Technical Specifications. Removal of the discussion of the NRC environmental monitoring program with the State reflects the cancellation of that program with the State. It does not alter any other environmental monitoring requirements. Therefore, the changes do not affect accident or transient initiation or consequences. As described above, the proposed changes are administrative in nature and do not impact the operation of any equipment needed for the mitigation of an accident or any known accident initiators.

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 changes are administrative in nature and therefore, do not change the fundamental requirements of the Technical Specifications. The proposed changes would not require any new or different accidents to be postulated, since no changes are being made to the plant that would introduce any new accident causal mechanisms. This license amendment request does not impact any plant systems that are potential accident initiators; nor does it have any significantly adverse impact on any accident mitigating systems.

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.

Since the proposed changes are administrative in nature, they do not change the fundamental requirements of the Technical Specifications. The proposed changes do not alter the permanent plant design, including instrument set points, nor does it change the assumptions contained in the safety analyses. Removal of the discussion of the NRC environmental monitoring program with the State reflects the cancellation of that program with the State. It does not alter any other environmental monitoring requirements.

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

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Acting Branch Chief: Douglas A. Broaddus.

Previously Published Notices 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 following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

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

Date of amendment request: February 12, 2013.

Brief description of amendment request: The proposed amendment would modify Cooper Nuclear Station license condition 2.E to require incorporation of the commitments listed in Appendix A of NUREG-1944, “Safety Evaluation Report Related to the License Renewal of Cooper Nuclear Station,” in the updated safety analysis report (USAR) to be managed in accordance with 10 CFR 50.59.

Date of publication of individual notice in Federal Register: July 5, 2013 (78 FR 40519).

Expiration date of individual notice: August 5, 2013 (public comments); September 3, 2013 (hearing requests).

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 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 [e-mail].

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: July 2, 2012, as supplemented by letters dated March 6 and May 28, 2013.

Brief description of amendments: The amendments revise Technical Specification (TS) 5.5.16 “Containment Leakage Rate Testing Program” by increasing the peak calculated containment internal pressure (P_a) from 49.4 pounds per square inch gauge (psig) to 49.7 psig for the design basis loss-of-coolant accident. In support of the revised P_a, the amendments also revise TS 3.6.4 “Containment Pressure” by decreasing the upper bound internal containment pressure limit from 1.8 psig to 1.0 psig.

Date of issuance: July 31, 2013.

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

Amendment Nos.: 303 and 281.

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

Date of initial notice in Federal Register: September 4, 2012 (77 FR 53926). The supplements dated March 6 and May 28, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated July 31, 2013.

No significant hazards consideration comments received: No.

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water Reactor, Vernon County, Wisconsin

Date of application for amendment: December 10, 2012, and supplemented February 25, 2013.

Brief description of amendment: The amendment revises the La Crosse Boiling Water Reactor License and Technical Specifications, as a result of the completion of the transfer of the spent fuel to dry cask storage.

Date of issuance: July 31, 2013.

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

Amendment No.: 72.

Facility Operating License No. DPR-7: This amendment revises the License and Technical Specifications.

Date of initial notice in Federal Register: March 19, 2013 (78 FR 16879).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 31, 2013.

No significant hazards consideration comments received: No.

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

Date of application of amendments: October 5, 2012.

Brief description of amendments: The amendments revised the Technical Specifications related to the integrated leak rate test of the reactor containment buildings.

Date of Issuance: August 5, 2013.

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

Amendment Nos.: 383 and 382.

Renewed Facility Operating License Nos. DPR-47 and DPR-55: Amendments revised the license and the technical specifications.

Date of initial notice in Federal Register: December 11, 2012, 77 FR 73688.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 5, 2013.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of application for amendment: November 9, 2012, as supplemented by letter dated on January 30, 2013.

Description of amendment: The amendment revised the Technical Specifications (TSs) to support the correction of a non-conservative TS allowable value in TS Table 3.3.6.1-1, “Allowable Value for Primary Containment and Drywell Isolation Instrumentation,” Function 3.c, “Reactor Core Isolation Cooling (RCIC) Steam Supply Line Pressure—Low.” This TS allowable value is changed from greater than or equal to 53 pounds per square inch (psig) to greater than or equal to 57 psig.

Date of issuance: August 5, 2013.

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

Amendment No: 194.

Facility Operating License No. NPF-29: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: February 5, 2013 (78 FR 8200). The supplemental letter dated January 30, 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 August 5, 2013.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 21, 2012.

Brief description of amendment: The amendment adopts NRC-approved Technical Specification Task Force (TSTF)—522, “Revise Ventilation System Surveillance Requirements to Operate For 10 Hours Per Month.” The amendment revises the Surveillance Requirement (SR) which currently requires operating the Standby Gas Treatment (SGT) System, with the electrical heaters operating, for a continuous 10 hour period at a frequency specified in the Surveillance Frequency Control Program. This Surveillance Requirement (SR 3.6.4.3.1) is revised to require operation of the system for 15 continuous minutes without the heaters operating.

In addition, the requirements for testing the SGT System specified in the Ventilation Filter Testing Program (VFTP) in Section 5.5.7, are revised accordingly to remove the electric heater output test (Specification 5.5.7.e) and to increase the specified relative humidity (RH) for the charcoal testing from the current 70% to 95% RH in Specification 5.5.7.c.

Date of issuance: July 25, 2013.

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

Amendment No.: 285.

Renewed Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 16, 2013 (78 FR 22571).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 25, 2013.

No significant hazards consideration comments received: No.

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

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

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

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

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

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

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

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

For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License or Combined License, as applicable, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items 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's 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 NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to [e-mail].

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC's Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. 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 petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. 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 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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

All documents filed in the NRC's 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 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 [e-mail], 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 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 NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

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

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with the NRC 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 to [e-mail], 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.

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

Date of amendment request: July 21, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No.ML13203A136), as supplemented by letter dated July 24, 2013 ADAMS Accession No. ML13206A042).

Description of amendment request: The amendment revised the Updated Safety Analysis Report (USAR) for the design basis tornado and tornado missiles to include Regulatory Guide 1.76, Revision 1, “Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants,” and Bechtel Power Corporation, Topical Report BC-TOP-9A, Revision 2, September 1974, “Design of Structures for Missile Impact.” The changes revise the current licensing basis pertaining to protection from tornadoes and tornado-generated missiles. RG 1.76, Revision 1 provides guidance for licensees to use in selecting the DBT and DBT-generated missiles that a nuclear power plant should be designed to withstand to prevent undue risk to public health and safety. BC-TOP-9A, Revision 2 provides a methodology for evaluating the impact of tornado missiles. The changes provide a means to analyze and document that the plant will be able to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by a tornado.

Date of issuance: July 26, 2013.

Effective date: As of its issuance date and shall be implemented upon approval.

Amendment No.: 272.

Renewed Facility Operating License No. DPR-40: The amendment revised the facility operating license.

Public comments requested as to proposed no significant hazards consideration: Yes (Omaha-World Herald, located in Omaha, Nebraska, on July 24 and 25, 2013). The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. One comment was received and evaluated.

The supplemental letter dated July 24, 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 Omaha-World Herald on July 24 and 25, 2013.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination (including the comment received on the NSHC) are contained in a safety evaluation dated July 26, 2013 (ADAMS Accession No. ML13203A070).

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

NRC Branch Chief: Michael T. Markley.

Dated at Rockville, Maryland, this 12th day of August 2013.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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