2013-06164. 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 (the Commission or 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 February 21, 2013, to March 6, 2013. The last biweekly notice was published on March 4, 2013 (78 FR 14126).

ADDRESSES:

You may access information and comment submissions related to this document, which the NRC possesses and is publically available, by searching on http://www.regulations.gov under Docket ID NRC-2013-0049. You may submit comments by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0049. 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: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
• Fax comments to: RADB at 301-492-3446.

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-2013-0049 when contacting the NRC about the availability of information regarding this document. You may access information related to this document by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0049.
• 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-2013-0049 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 that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. 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 “Rules of Practice for Domestic Licensing Proceedings” 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 NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC 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 the NRC guidance available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the agency's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC's Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by email at [e-mail], or by a toll-free call at 1-866ndash;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) The information upon which the filing is based was not previously available; (ii) the information upon which the filing is based is materially different from information previously available; and (iii) the filing has been submitted in a timely fashion based on the availability of the subsequent information.

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail].

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

Date of amendment request: December 10, 2012.

Description of amendment request: The proposed amendment would revise certain license conditions and to remove TS definitions, operational requirements, and specific design requirements that are no longer applicable with all spent fuel in dry cask storage at the Independent Spent Fuel Storage Installation (ISFSI). The proposed changes to the TS also remove administrative control requirements that have been relocated to the LACBWR Quality Assurance Program Description (QAPD) or are superseded by regulation or other guidance.

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

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

Response: No.

The proposed changes reflect the complete transfer of all spent nuclear fuel from the Fuel Element Storage Well (FESW) to the Independent Spent Fuel Storage Installation (ISFSI). Design basis SAFSTOR accidents related to the FESW were discussed in the LACBWR Decommissioning Plan. These postulated accidents were predicated on spent nuclear fuel being stored in the FESW. With the removal of the spent fuel from the FESW, there are no remaining important to safety systems required to be monitored and there are no remaining credible accidents that require that actions of a Certified Fuel Handler to prevent occurrence or mitigate the consequences.

The LACBWR Decommissioning Plan provided a discussion of radiological events postulated to occur during SAFSTOR with the bounding consequence resulting from a materials handling event. The proposed changes do not have an adverse impact on decommissioning activities or any postulated consequences.

The proposed change to the Design Features section of the Technical Specifications clarifies that the spent fuel is being stored in dry casks within an ISFSI. The probability or consequences of accidents at the ISFSI are evaluated in the dry cask vendor's FSAR and are independent of the SAFSTOR accidents that were evaluated in the LACBWR Decommissioning Plan.

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

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

Response: No.

The proposed changes reflect the reduced operational risks as a result of the spent nuclear fuel being transferred to dry casks within an ISFSI. The proposed changes do not modify any physical systems, or components. The plant conditions for which the LACBWR Decommissioning Plan design basis accidents relating to spent fuel were evaluated are no longer applicable. The proposed changes do not affect any of the parameters or conditions that could contribute to the initiation of an accident. Design basis accidents associated with the dry cask storage of spent fuel are already considered in the dry cask system's Final Safety Analysis Report. No new accident scenarios are created as a result of deleting non-applicable operational and administrative requirements.

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

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

Response: No.

As described above, the proposed changes reflect the reduced operational risks as a result of the spent nuclear fuel being transferred to dry casks within an ISFSI. The design basis and accident assumptions within the LACBWR Decommissioning Plan and the Technical Specifications relating to spent fuel are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities. In addition, the proposed changes do not result in a change in initial conditions, system response time, or in any other parameter affecting the SAFSTOR accident analysis.

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: Mr. Thomas Zaremba, Wheeler, Van Sickle and Anderson, Suite 801, 25 West Main Street, Madison, WI 53703-3398.

NRC Branch Chief: Bruce Watson.

Detroit Edison, Docket No. 50-016, Fermi 1, Monroe County, Michigan

Date of amendment request: December 21, 2012.

Description of amendment request: The proposed amendment (ML13002A037) would revise the Fermi 1 operating license to change its name on the license to “DTE Electric Company.” This name change is purely administrative in nature. Detroit Edison is a wholly owned subsidiary of DTE Energy Company, and this name change is part of a set of name changes of DTE Energy subsidiaries to conform their names to the “DTE” brand name. No other changes are contained within this request. This request does not involve a transfer of control over or of an interest in the license for Fermi 1.

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

The proposed amendment changes the name of the owner licensee. The proposed amendment is purely administrative in nature. The functions, powers, resources and management of the owner licensee will not change. Detroit Edison, which will be renamed DTE Electric Company, will remain the licensee of the facility. The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

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

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

The proposed amendment is purely administrative in nature. The functions of the owner licensee will not change. These changes do not involve any physical alteration of the plant (i.e., no new or different type of equipment will be installed), and installed equipment is not being operated in a new or different manner. Thus, no new failure modes are introduced.

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

3. The proposed change does not involve a significant reduction in the margin of safety.

The proposed amendment is a name change to reflect the new name of the owner licensee. The proposed amendment is purely administrative in nature. The functions of the owner licensee will not change. Detroit Edison, which will be renamed DTE Electric Company, will remain the licensee of the facility, and its functions will not change. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There are no changes to setpoints at which protective actions are initiated, and the operability requirements for equipment assumed to operate for accident mitigation are not affected.

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

NRC Branch Chief: Bruce Watson.

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

Date of amendment request: January 11, 2013.

Description of amendment request: The proposed amendment would update the Fermi 2 Updated Final Safety Analysis Report (UFSAR) to describe methodology and results of the analysis performed to evaluate the protection of the plant's structures, systems and components (SSCs) from tornado generated missiles. The analysis is consistent with the guidance provided in Regulatory Issue Summary 2008-14, “Use of TORMIS Computer Code for Assessment of Tornado Missile Protection.”

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.

Proposed for NRC review and approval are changes to the Fermi 2 Updated Final Safety Analysis Report (UFSAR) which in essence constitute a license amendment to incorporate use of an NRC approved methodology to assess the need for additional positive (physical) tornado missile protection of specific features at the Fermi 2 site. The UFSAR changes will reflect use of the Electric Power Research Institute (EPRI) Topical Report “Tornado Missile Risk Evaluation Methodology” (EPRI NP-2005), Volumes I and II. As noted in the NRC Safety Evaluation Report on this topic dated October 26, 1983, the current licensing criteria governing tornado missile protection are contained in Standard Review Plan (SRP) Sections 3.5.1.4 and 3.5.2. These criteria generally specify that safety-related systems be provided positive tornado missile protection (barriers) from the maximum credible tornado threat. However, SRP Section 3.5.1.4 includes acceptance criteria permitting relaxation of the above deterministic guidance, if it can be demonstrated that the probability of damage to unprotected essential safety-related features is sufficiently small.

As permitted in NRC Standard Review Plan (NUREG-0800) sections, the combined probability will be maintained below an allowable level, i.e., an acceptance criterion threshold, which reflects an extremely low probability of occurrence. The Fermi 2 approach assumes that if the sum of the individual probabilities calculated for tornado missiles striking and damaging portions of important systems or components is greater than or equal to 10⁻⁶ per year per unit, then installation of unique missile barriers would be needed to lower the total cumulative probability below the acceptance criterion of 10⁻⁶ per year per unit.

With respect to the probability of occurrence or the consequences of an accident previously evaluated in the UFSAR, the possibility of a tornado reaching the Fermi 2 site and causing damage to plant structures, systems and components is a design basis event considered in the Updated Final Safety Analysis Report. The changes being proposed do not affect the probability that the natural phenomenon (a tornado) will reach the plant, but from a licensing basis perspective they do affect the probability that missiles generated by the winds of the tornado might strike and damage certain plant systems or components. There are a limited number of safety-related components that could theoretically be struck and consequently damaged by tornado-generated missiles. The probability of tornado-generated missile strikes on “important” systems and components (as discussed in Regulatory Guide 1.117, “Tornado Design Classification”) is what is to be analyzed using the probability methods discussed above. The combined probability of damage will be maintained below an extremely low acceptance criterion to ensure overall plant safety. The proposed change is not considered to constitute a significant increase in the probability of occurrence or the consequences of an accident, due to the extremely low probability of damage due to tornado-generated missiles and thus an extremely low probability of a radiological release.

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

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

Response: No.

The possibility of a tornado reaching the Fermi 2 site is a design basis event that is explicitly considered in the UFSAR. This change involves recognition of the acceptability of performing tornado missile probability calculations in accordance with established regulatory guidance. The change therefore deals with an established design basis event (the tornado). Therefore, the proposed change would not contribute to the possibility of a new or different kind of accident from those previously analyzed. The probability and consequences of such a design basis event are addressed in Question 1 above.

Based on the above discussions, the proposed change will not create the possibility of a new or different kind of accident than those previously evaluated.

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

Response: No.

The existing Fermi 2 licensing basis for protection of safety-related equipment required for safe shutdown from design basis tornado generated missiles is to provide positive missile barriers for all safety-related systems and components. With the change, it will be recognized that there is an extremely low probability, below an established acceptance limit, that a limited subset of the “important” systems and components could be struck and consequently damaged. The change from protecting all safety-related systems and components to ensuring an extremely low probability of occurrence of tornado-generated missile strikes and consequential damage on portions of important systems and components is not considered to constitute a significant decrease in the margin of safety due to that extremely low probability.

Therefore, the changes associated with this license amendment request 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: Bruce R. Masters, DTE Energy, General Council—Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: December 17, 2012.

Description of amendment request: The proposed amendment would revise VYNPS Technical Specification (TS) 3.3.B to provide an action statement for inoperable control rods consistent with the Standard Technical Specification (STS) provision (NUREG-1433, Revision 4).

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 amendment does not significantly increase the probability or consequences of an accident. The adding of an additional, restrictive action statement for inoperable equipment, consistent with the STS does not alter any accident analysis.

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

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

Response: No.

The proposed amendment does not involve any new modes of operation. The change establishes additional restrictive controls for equipment that is considered inoperable. The proposed amendment does not change how the control rod system is operated or change the design configuration of the control rods. No new accident precursors are introduced. No new or different types of equipment will be installed. The methods governing plant operation remain bounded by current safety analysis assumptions.

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

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

Response: No.

The proposed amendment does not involve any new methods of operation. The change establishes additional restrictive controls for equipment that is considered inoperable. The proposed amendment does not change how the control rod system is operated or change the design configuration of the control rods. No new or different types of equipment will be installed. The methods governing plant operation remain bounded by current safety analysis assumptions.

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

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

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: George Wilson.

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

Date of amendment request: December 21, 2012.

Description of amendment request: The proposed amendment would revise the licensing basis relative to how the station satisfies the requirements in 10 CFR 50.63, “Loss of all alternating current power.” The VYNPS currently relies on the Vernon Hydroelectric Station (VHS) as the alternate alternating current (AAC) power source providing acceptable capability to withstand station blackout under 10 CFR 50.63(c)(2). The VYNPS proposes to replace the VHS with an onsite diesel generator as the AAC power source providing this capability which would involve changes to the facility and procedures described in the VYNPS Updated Final Safety Analysis 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment does not significantly increase the probability or consequences of an accident. The proposed amendment replaces one AAC power source (the VHS) with an additional onsite AAC power source (diesel generator). This equipment can not initiate a design basis accident and is not used to mitigate the consequences of design basis accidents. The equipment is used to mitigate the consequences of a station blackout as required by 10 CFR 50.63. Station blackout events are not considered design basis accidents and do not result in radiological consequences.

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

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

Response: No.

The proposed amendment does not involve any new modes of operation. The change provides an alternate means to provide AAC power to the station. The location of the SBO DG does not create the possibility of a different kind of accident. No new accident precursors are introduced. Station procedures will be revised to align the AAC source to provide the required power within established coping times. The methods governing plant operation remain bounded by current safety analysis assumptions.

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

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

Response: No.

The design of the new AAC source will accommodate the loading associated with the proceduralized station blackout response and safety margins will be maintained. The design of the system will meet regulatory guidance and be within station design analysis. The station safety analysis results are unchanged and margin to regulatory limits is not affected.

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

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

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: George Wilson.

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

Date of amendment request: December 14, 2012, as supplemented by letter dated January 31, 2013.

Description of amendment request: The proposed amendment would modify the pressure-temperature limit curves and low temperature overpressure protection limits in the Three Mile Island Nuclear Station, Unit 1 Technical Specification (TS) Section 3.1.2, “Pressurization Heatup and Cooldown Limitations,” TS Section 3.1.12, “Pressurizer Power Operated Relief Valve, Block Valve, and Low-Temperature Overpressure Protection,” and TS Section 4.5.2, “Emergency Core Cooling System.” The proposed changes reflect revised fluence projections out to 50.2 effective full-power years (EFPY) as compared to the current projections which go to 29 EFPY. The submittal, dated December 14, 2012, also includes a corresponding exemption request to use an alternate initial reference temperature for nil-ductility transition (RT_NDT) for Linde 80 weld materials per NRC-approved Topical Report BAW-2308, “Initial RT_NDT of Linde 80 Weld Materials,” Revisions 1-A and 2-A.

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.

The proposed amendment will revise the reactor coolant system heatup, cooldown, and inservice leak hydrostatic test limitations (Technical Specification (TS) Section 3.1.2 (“Pressurization Heatup and Cooldown Limitations”)) for the Reactor Coolant System (RCS) to a maximum of 50.2 Effective Full Power Years (EFPY) in accordance with 10 CFR Part 50, Appendix G. Further, the proposed amendment revises TMI, Unit 1 Technical Specification Sections 3.1.12 (“Pressurizer Power Operated Relief Valve (PORV), Block Valve, and Low Temperature Overpressure Protection (LTOP)”), and 4.5.2 (“Emergency Core Cooling System”) for Low Temperature Overpressure Protection (LTOP) requirements to reflect the revised P-T limits of the reactor vessel. P-T limits for the TMI, Unit 1 reactor vessel were developed in accordance with the requirements of 10 CFR Part 50, Appendix G (“Fracture Toughness Requirements”), utilizing the analytical methods and flaw acceptance criteria of Topical Report BAW-10046A (AREVA NP Document BAW-10046A, Rev. 2, “Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR Part 50, Appendix G,” by H. W. Behnke et al., June 1986) and ASME Code Section XI, Appendix G (“Fracture Toughness Criteria for Protection Against Failure,” 1995 Edition with Addenda through 1996) which are previously approved NRC standards for the preparation of P-T limit curves. Updating the P-T limit curves for additional EFPY maintains the level of assurance that Reactor Coolant Pressure Boundary integrity will be maintained, as specified in 10 CFR Part 50, Appendix G. Additionally, this proposed amendment deletes administrative requirements contained in TS 3.1.2.4 and 3.1.2.5 which provide reporting requirements related to the preparation and submittal of P-T curves that are outdated or contained in regulation.

The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

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

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

Response: No.

The proposed changes incorporate methodologies that either have been approved or accepted for use by the NRC (provided that any conditions/limitations are satisfied). The P-T limit curves and LTOP limits will provide the same level of protection to the Reactor Coolant Pressure Boundary as was previously evaluated. Reactor Coolant Pressure Boundary integrity will continue to be maintained in accordance with 10 CFR Part 50, Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. Additionally, this proposed amendment deletes administrative requirements contained in TS 3.1.2.4 and 3.1.2.5. These changes do not involve any physical alteration of the plant (i.e., no new or different type of equipment will be installed), and installed equipment is not being operated in a new or different manner. Thus, no new failure modes are introduced.

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

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

Response: No.

The proposed changes do not affect the function of the Reactor Coolant Pressure Boundary or its response during plant transients. By calculating the P-T limits and associated LTOP limits using NRC-approved methodology, adequate margins of safety relating to Reactor Coolant Pressure Boundary integrity are maintained. Additionally, this proposed amendment deletes administrative requirements contained in TS 3.1.2.4 and 3.1.2.5. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. These changes will ensure that protective actions are initiated and the operability requirements for equipment assumed to operate for accident mitigation are not affected.

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: J. Bradley Fewell, Esquire, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Meena Khanna.

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

Date of amendment request: February 4, 2013.

Description of amendment request: The proposed amendment would delete various reporting requirements contained in the Technical Specifications (TSs). Specifically, the proposed amendment will delete the Sealed Source Contamination Special Report and the Startup Report, as well as the plant-specific annual reports regarding periodic Leak Reduction Program tests, Pressurizer Power Operated Relief Valve and Pressurizer Safety Valve challenges, specific activity analysis in which the primary coolant exceeds the limits of TS 3.1.4.1, and major changes to radioactive waste treatment systems.

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

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

Response: No.

The proposed changes do not involve the modification of any plant equipment or affect plant operation. The proposed changes will have no impact on any safety related structures, systems, or components. The reporting requirements proposed for deletion are not required because the requirements are adequately addressed by other regulatory requirements, or are no longer warranted.

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

Response: No.

The proposed changes have no impact on the design, function or operation of any plant structure, system or component. The proposed changes do not affect plant equipment or accident analyses. The reporting requirements proposed for deletion are not required because the requirements are adequately addressed by other regulatory requirements, or are no longer warranted.

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

Response: No.

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

Margins of safety are unaffected by deletion of the reporting requirements.

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

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

NRC Branch Chief: Meena Khanna.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1 (DBNPS), Ottawa County, Ohio

Date of amendment request: January 18, 2013.

Description of amendment request: The amendment would revise DBNPS Technical Specification (TS) 3.4.17, “Steam Generator (SG) Tube Integrity”; TS 3.7.18, “Steam Generator Level”; TS 5.5.8, “Steam Generator (SG) Program”; and TS 5.6.6, “Steam Generator Tube Inspection Report.” The proposed revision to these TSs is to support plant operations following the replacement of the original SGs which is scheduled to be completed in April 2014. The proposed changes to TS 3.4.17, TS 5.5.8, and TS 5.6.6 would impose requirements that reflect the analysis and tube materials of the replacement SGs. These changes are consistent with Technical Specifications Task Force (TSTF) traveler TSTF-510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,” which was approved by the U.S. Nuclear Regulatory Commission on October 27, 2011. The proposed revision to TS 5.5.8 also includes minor editorial changes and eliminates the requirements for special visual inspections of the internal auxiliary feedwater header, since this component will not be part of the replacement SGs.

The proposed changes to TS 3.7.18 would impose inventory limits on the secondary-side that reflect the design characteristics and dimensions of the replacement SGs. The revised limits will ensure that plant operations with the replacement SGs is bounded by the values used in the existing main steam line break analysis presented in the DBNPS updated safety analysis 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 consequences of an accident previously evaluated?

Response: No.

For TS 3.4.17, “Steam Generator (SG) Tube Integrity,” a steam generator tube rupture (SGTR) event is the relevant design basis accident analyzed in the licensing basis for DBNPS. TS 3.4.17 and TS 5.5.8, “Steam Generator (SG) Program,” impose monitoring and inspection requirements that ensure tube integrity is maintained. The proposed changes to these TSs would implement monitoring and inspection requirements appropriate for the design and materials of the replacement SGs. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that that the integrity of the SG tubes is verified to be maintained at a level that prevents an increase in the probability of a SGTR.

Therefore the proposed changes to these TSs will not increase the probability of a SGTR.

The radiological consequences of a SGTR are bounded by using conservative assumptions in the design basis accident analysis, and are dependent upon the pre-existing primary-to-secondary leak rate, the flow rate through the ruptured tube, the radiological isotopic content of the RCS [reactor coolant system] and the release paths. The monitoring and inspection requirements imposed by TS 3.4.17 and TS 5.5.8 are intended to ensure that SG tube integrity is maintained. The proposed changes to these TSs would implement monitoring and inspection requirements appropriate for the design and materials of the replacement SGs and would not affect radiological releases in the event of an SGTR. The radiological isotopic content of the RCS and the release paths are not affected by any of the requirements in the current TS 3.4.17 or TS 5.5.8 or proposed revisions thereto. Therefore, the proposed changes to these TSs will not increase the consequences of a SGTR.

TS 5.6.6, “Steam Generator Tube Inspection Report,” specifies information that is to be reported to the NRC following SG inspections performed in accordance with the Steam Generator Program requirements contained in TS 5.5.8. The requirement to provide this report is administrative in nature and the content of this report can have no effect on the probability or the consequences of an accident previously evaluated.

LCO [limiting condition for operation] 3.7.18, “Steam Generator Level” ensures that the plant is operated within the SG inventory limits that were used as initial conditions in the current accident analysis for a Main Steam Line Break (MSLB). The SG inventory is not an accident initiator and does not affect any accident initiator. Therefore, the proposed changes in SG inventory limits will not increase the probability of a MSLB accident.

The radiological consequences of a MSLB are dependent upon the total SG inventory released, the SG primary-to-secondary leakage rate, the radiological isotopic content of the RCS, and the release paths. The revision to LCO 3.7.18 will ensure that the total inventory released remains bounded by the existing analysis. None of the other factors listed above are affected by the revised operating limits on SG inventory that are proposed in the revisions to LCO 3.7.18.

Therefore, the proposed changes in SG inventory limits will not increase the consequences of a MSLB.

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

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

Response: No.

The proposed changes support replacement of the SGs at the DBNPS. Replacement of the SGs is being performed as a design modification in accordance with the provisions of 10 CFR 50.59, “Changes, tests and experiments.” The proposed changes to TS 3.4.17, TS 5.5.8 and TS 5.6.6 would implement monitoring and inspection requirements appropriate for the design and materials of the replacement SGs, and establish appropriate reporting requirements. These changes would not affect the method of operation of the SGs. The proposed changes to TS 3.7.18 would ensure that the replacement SGs will be operated in accordance with existing analyses. None of the proposed changes would introduce any changes to the plant design. In addition, the proposed changes would not impact any other plant system or component.

The proposed changes would continue to prevent loss of SG tube integrity, and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by these changes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

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

The proposed changes will ensure that the existing margins of safety are maintained following the replacement of SGs. The changes to LCO 3.4.17 and TSs 5.5.8 and 5.6.6 impose requirements for SG tube integrity monitoring, inspection, and reporting that will ensure that there is no reduction in the ability of the tubes to perform their RCS pressure boundary and heat transfer functions. The changes to LCO 3.7.18 ensure the MSLB accident analyses remain bounding.

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: David W. Jenkins, FirstEnergy Corporation, 76 South Main Street, Akron, Ohio 44308.

NRC Branch Chief: Jeremy S. Bowen.

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

Date of amendment request: December 27, 2012.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to align St. Lucie TSs with Combustion Engineering Owners Group TSs language describing required licensed Senior Reactor Operator (SRO) duties during fuel handling activities.

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

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

Response: No.

The proposed changes will not result in any significant increase in the probability or consequences of an accident previously evaluated, as the proposed TS changes are consistent with Standard Technical Specifications. Further, not requiring licensed SRO oversight of fuel handling operations other than core alterations does not introduce additional risk or a greater potential for consequences of an accident that has not previously been evaluated.

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

Response: No.

The proposed changes are administrative in nature and do not involve a physical modification of the plant. No new or different type of equipment will be installed. The methods for conducting core alterations and other fuel handling operations will remain the same. The proposed changes will not introduce new failure modes/effects that could lead to an accident for which consequences exceed that of accidents previously analyzed. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes will not involve a significant reduction in a margin of safety in that the changes are administrative in nature. No plant equipment or accident analyses will be affected. Additionally, the proposed changes will not relax any criteria used to establish safety limits, safety system settings, or the bases for any limiting conditions for operation. Safety analysis acceptance criteria are not affected. Plant operation will continue within the design basis. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and maintain the plant in a safe shutdown condition. Consequently, the proposed changes do not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Jessie F. Quichocho.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: June 6, 2012.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to eliminate the requirements that the average power range monitoring (APRM) system “Upscale” and “Inoperative” scram and control rod withdrawal block functions be operable in Operational Condition (OPCON) 5, refueling operations.

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 with the NRC staff's edits in square brackets:

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

Response : No.

The APRM system is not an initiator of or a precursor to any accident or transient. The APRM system monitors the neutron flux level in the power operating range from approximately one percent to greater than rated thermal power and initiates automatic protective actions for postulated at-power reactivity insertion events. Thus, the proposed changes to the TS operability requirements for the APRM system will not impact the probability of any previously evaluated accident.

The design of plant equipment is not being modified by the proposed amendment. The TSs will continue to require operability of the APRM system “Upscale” and “Inoperative” scram and control rod withdrawal block functions when the reactor is in the Startup and Run modes (OPCON 2 and OPCON 1) to provide core protection for postulated reactivity insertion events occurring during power operating conditions. Thus, the consequences of previously evaluated at-power reactivity insertion events are not affected by the proposed amendment.

The proposed elimination of the TS requirements that the APRM system “Upscale” and “Inoperative” scram and control rod withdrawal block functions be operable when the reactor is in the Refueling mode (OPCON 5) also does not increase the consequences of an accident previously evaluated. The possibility of inadvertent criticality due to a control rod withdrawal error during refueling is minimized by design features and procedural controls that are not affected by the proposed amendment. Since the core is designed to meet shutdown requirements with the highest worth rod withdrawn, the core remains subcritical even with one rod withdrawn. Any attempt to withdraw a second rod results in a rod block by the Refueling Interlocks (RI). In addition, since reactor neutron flux levels during refueling are below the APRM indicating range, the APRM system does not provide any meaningful core monitoring or protection in the refueling operating condition (OPCON 5). The source range (SRM) and intermediate range (IRM) neutron monitoring systems provide adequate neutron flux monitoring during refueling and automatically initiate protective actions (scram or control rod withdrawal block) when required during refueling.

Additionally, if the infrequently performed TS 3/4.10.3, “Shutdown Margin Demonstrations,” is performed in OPCON 5, the additional controls and restrictions in place during this test are sufficiently robust even without the RIs when the mode switch is temporarily placed in Startup. In addition to the OPCON 5 SRM and IRM protective actions, the SRM RPS [reactor protection system] trip is made operable, the RWM [rod worth minimizer] is operable and programmed for the shutdown margin demonstration, use of the “rod-out-notch-override” control is prohibited, and no other core alterations are allowed. Therefore, during this infrequent operation, operability of the APRMs is not required as they would not provide any meaningful core monitoring or protection.

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 to the TS operability requirements for the APRM system do not introduce any new accident precursors and do not involve any physical plant alterations or changes in the methods governing normal plant operation that could initiate a new or different kind of accident. The proposed amendment does not alter the intended function of the APRM system and does not affect the ability of the system to provide core protection for at-power reactivity insertion events. The other existing TS-required neutron monitoring systems (SRM and IRM) provide for core monitoring and protection in the refueling mode (OPCON 5). Additionally, if the infrequently performed TS 3/4.10.3, “Shutdown Margin Demonstrations” is performed in OPCON 5, the additional controls and restrictions in place during this test are sufficiently robust even without the RIs when the mode switch is temporarily placed in “Startup.”

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

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

Response: No.

Margin of safety is related to the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed amendment does not alter setpoints or limits established or assumed by the accident analyses. The proposed TS changes to eliminate the requirements that the APRM system “Upscale” and “Inoperative” scram and control rod withdrawal block functions be operable when in OPCON 5 have no impact on the performance of the fission product barriers. These APRM functions do not provide any meaningful core monitoring or protection in the Refueling operating condition, including the infrequently performed special test TS 3/4.10.3. The other existing TS required neutron monitoring systems (SRM and IRM) provide for core monitoring and protection in the refueling mode (OPCON 5). In the Startup and Run modes the TSs will continue to require operability of these APRM functions to provide core protection for postulated reactivity insertion events occurring during power operating conditions, consistent with the plant safety analyses.

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, and with the changes noted above in square brackets, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Branch Chief: Meena K. Khanna.

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

Date of amendment request: February 15, 2013.

Description of amendment request: The proposed change would amend Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric Generating Plant (VEGP) Units 3 and 4 by departing from the plant-specific design control document Tier 2* material by revising reference document APP-OCS-GEH-320, “AP1000 Human Factors Engineering Integrated System Validation Plan” from Revision D to Revision 2. APP-OCS-GEH-320 is incorporated by reference in the updated final safety analysis report (UFSAR) as a means to implement the activities associated with the human factors engineering verification and validation.

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 Integrated System Validation (ISV) provides a comprehensive human performance-based assessment of the design of the AP1000 Human-System Interface (HSI) resources, based on their realistic operation within a simulator-driven Main Control Room (MCR). The ISV is part of the overall AP1000 Human Factors Engineering (HFE) program. The changes are to the ISV Plan to clarify the scope and amend the details of the methodology. The ISV Plan is needed to perform, in the simulator, the scenarios described in the document. The functions and tasks allocated to plant personnel can still be accomplished after the proposed changes. The performance of the tests governed by the ISV Plan provides additional assurances that the operators can appropriately respond to plant transients. The ISV Plan does not affect the plant itself. Changing the ISV Plan does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related structure, system, component (SSC) or function is adversely affected. The changes do not involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. Because the changes do not involve any safety-related SSC or function used to mitigate an accident, the consequences of the accidents evaluated in the UFSAR are not affected.

Therefore, there is no significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The changes to the ISV Plan affect the testing and validation of the Main Control Room and Human System Interface using a plant simulator.

Therefore, the changes do not affect the safety-related equipment itself, nor do they affect equipment which, if it failed, could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected. No system or design function or equipment qualification will be adversely affected by the changes. This activity will not allow for a new fission product release path, nor will it result in a new fission product barrier failure mode, nor create a new sequence of events that would result in significant fuel cladding failures. In addition, the changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment.

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

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

Response: No.

The changes to the ISV Plan affect the testing and validation of the Main Control Room and Human System Interface using a plant simulator. Therefore, the changes do not affect the assessments or the plant itself. These changes do not affect safety-related equipment or equipment whose failure could initiate an accident, nor does it adversely interface with safety-related equipment or fission product barriers. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change.

Therefore, there is no significant reduction in a margin of safety.

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

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

NRC Acting Branch Chief: Lawrence Burkhart.

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

Date of amendment request: November 19, 2012.

Description of amendment request: The proposed amendment would change the Technical Specification (TS) 3.7.10 to require a unit shutdown within the TS 3.7.10 Actions instead of entering Limiting Condition for Operation (LCO) 3.0.3 when both Control Room Emergency Ventilation System (CREVS) trains are inoperable in MODE 1, 2, 3, or 4 due to actions taken as a result of a tornado warning and the Completion Time of 8 hours for restoration of at least one CREVS train to OPERABLE status is not met.

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 modify WBN Unit 1 TS 3.7.10 to resolve a potential conflict in applying the appropriate actions for not meeting the Required Action and associated Completion Time of Condition E. These proposed changes are acceptable in the event that both CREVS trains are inoperable in MODE 1, 2, 3, or 4 due to actions taken as a result of a tornado warning and the Completion Time of 8 hours for restoration of at least one CREVS train to OPERABLE status is not met because the requirements to shutdown the unit to Mode 3 and Mode 5 are similar to the current requirements, the required Completion Times are 1 hour less than the existing LCO 3.0.3 Completion Times that currently apply, and do not impact the design and operation of the CREVS, or the ultimate Actions required to be taken by TS 3.7.10 upon inoperability of the CREVS in MODE 1, 2, 3, or 4 due to actions taken as a result of a tornado warning. The proposed changes do not (1) require physical changes to plant systems, structures, or components; (2) prevent the safety function of any safety-related system, structure, or component during a design basis event; (3) alter, degrade, or prevent action described or assumed in any accident described in the WBN Unit 1 UFSAR from being performed since the safety-related systems, structures, or components are not modified; (4) alter any assumptions previously made in evaluating radiological consequences; or (5) affect the integrity of any fission product barrier.

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 modify WBN Unit 1 TS 3.7.10 to resolve a potential conflict in applying the appropriate, actions for not meeting the Required Action and associated Completion Time of Condition E. These proposed changes are acceptable in the event that both CREVS trains are inoperable in MODE 1, 2, 3, or 4 due to actions taken as a result of a tornado warning and the Completion Time of 8 hours for restoration of at least one CREVS train to OPERABLE status is not met because the requirements to shutdown the unit to Mode 3 and Mode 5 are similar to the current requirements, the required Completion Times are 1 hour less than the existing LCO 3.0.3 Completion Times that currently apply, and do not impact the design and operation of the CREVS, or the ultimate Actions required to be taken by TS 3.7.10 upon inoperability of the CREVS in MODE 1, 2, 3, or 4 due to actions taken as a result of a tornado warning. The proposed changes do not introduce any new accident causal mechanisms, since no physical changes are being made to the plant, nor do they impact any plant systems that are potential accident initiators.

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

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

Response: No.

The proposed changes modify WBN Unit 1 TS 3.7.10 to resolve a potential conflict in applying the appropriate actions for not meeting the Required Action and associated Completion Time of Condition E. These proposed changes are acceptable in the event that both CREVS trains are inoperable in MODE 1, 2, 3, or 4 due to actions taken as a result of a tornado warning and the Completion Time of 8 hours for restoration of at least one CREVS train to OPERABLE status is not met because the requirements to shutdown the unit to Mode 3 and Mode 5 are similar to the current requirements, the required Completion Times are 1 hour less than the existing LCO 3.0.3 Completion Times that currently apply, and do not impact the design and operation of the CREVS, or the ultimate Actions required to be taken by TS 3.7.10 upon inoperability of the CREVS in MODE 1, 2, 3, or 4 due to actions taken as a result of a tornado warning. As such, there is no impact on the safety analysis for the CREVS. The proposed changes do not alter the permanent plant design, including instrument set points, that is the basis of the assumptions contained in the safety analyses.

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 Branch Chief: Jessie F. Quichocho.

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 Commission'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].

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of application for amendments: March 6, 2012, as supplemented by letters dated August 29, 2012, September 21, 2012, November 29, 2012, and January 22, 2013.

Brief Description of amendments: The amendments revise Technical Specification (TS) 5.6.5.b by replacing AREVA Topical Report ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors with AREVA Topical Report ANP-I 0307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, June 2011, in the list of analytical methods that have been reviewed and approved by the U.S. Nuclear Regulatory Commission for determining core operating limits, (2) revise TS 2.1.1, “Reactor Core SLs [Safety Limits],” by incorporating revised Safety Limit Minimum Critical Power Ratio (SLMCPR) values, and (3) revise the license condition in Appendix B, “Additional Conditions,” of the operating licenses regarding an alternate method for evaluating SLMCPR values.

Date of issuance: March 1, 2013.

Effective date: Date of issuance, to be implemented prior to the startup from the 2014 Unit 1 refueling outage for Unit 1 changes, and prior to the startup from the 2013 Unit 2 refueling outage for Unit 2 changes.

Amendment Nos.: Unit 1—262 and Unit 2—290.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments change the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: July 3, 2012 (77 FR 39524). The supplements dated August 29, 2012, September 21, 2012, November 29, 2012, and January 22, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 1, 2013.

No significant hazards consideration comments received: No.

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

Date of application for amendment: August 6, 2012.

Brief Description of amendment: The amendment allows a delay time for entering a supported system Technical Specification (TS) when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.8 is added to the TS to provide this allowance and define the requirements and limitations for its use.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff issued a notice of opportunity for comment in the Federal Register on November 24, 2004 (69 FR 68412), on possible amendments concerning TSTF-372, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252).

Date of issuance: February 26, 2013.

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

Amendment No.: 232.

Renewed Facility Operating License No. DPR-23: Amendment changed the license and TSs.

Date of initial notice in Federal Register: October 16, 2012 (77 FR 63347).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 26, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 13, 2012.

Description of amendment request: The proposed amendment would revise the Millstone Power Station, Unit 2 (MPS2) Technical Specification (TS) requirements related to diesel fuel oil testing consistent with NUREG-1432, Rev. 3.1, “Standard Technical Specifications, Combustion Engineering Plants,” December 1, 1995, and NRC approved Technical Specification Task Force (TSTF) TSTF-374, “Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil,” Revision 0.

Date of issuance: March 5, 2013.

Effective date: As of the date of issuance, and shall be implemented within 120 days. Amendment No.: 313.

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

Date of initial notice in Federal Register: June 12, 2012 (77 FR 35072). The supplemental letter dated May 7, 2012, 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.

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: August 14, 2012, as supplemented by letters dated October 25, November 14, and December 13, 2012, and February 15, 2013.

Brief description of amendment: The amendment revises Technical Specification 3.5.4, “Refueling Water Storage Tank,” to permit non-seismically qualified piping of the Spent Fuel Pool purification system to be connected to the Refueling Water Storage Tank seismic piping under administrative controls for a limited period of time in order to purify the contents of the Refueling Water Storage Tank.

Date of issuance: February 22, 2013.

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

Amendment No.: 250.

Facility Operating License No. DPR-64: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: October 16, 2012 (77 FR 63350). The letters dated October 25, November 14, and December 13, 2012, and February 15, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 22, 2013.

No significant hazards consideration comments received: No.

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

Date of application for amendment: February 28, 2012, supplemented by letters dated September 6, 2012, November 7, 2012, November 29, 2012, February 21, 2013 and February 25, 2013.

Brief description of amendment: The amendment revises the PNP TSs to support the replacement of the Region I main spent fuel (SFP) storage racks and the storage racks in the north tilt pit portion of the SFP, with new neutron absorber Metamic-equipped racks. The replacement of the SFP storage racks will allow recovery of the currently unusable storage locations in the SFP.

Date of issuance: February 28, 2013.

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

Amendment No.: 250.

Facility Operating License No. DPR-20: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 5, 2012 (77 FR 33246). The supplemental letters dated September 6, 2012, November 7, 2012, November 29, 2012, February 21, 2013 and February 25, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 28, 2013.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-374, LaSalle County Station, Unit 2, LaSalle County, Illinois

Date of application for amendments: October 11, 2012, as supplemented by letters dated January 17, February 20, and February 26, 2013.

Brief description of amendments: The amendment request proposed changes to the Technical Specifications (TSs) to revise Section 2.1.1, “Reactor Core SLs,” minimum critical power ratio safety limit (MCPR SL) from ≥ 1.11 to ≥ 1.14 for two-loop recirculation operation and from ≥ 1.12 to ≥1.17 for a single-loop recirculation operation.

Date of issuance: February 27, 2013.

Effective date: As of the date of issuance and shall be implemented after Cycle 14 is completed and prior to the operation of Cycle 15.

Amendment No.: 192.

Facility Operating License Nos. NPF-18: The amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: November 5, 2012 (77 FR 66489).

The January 17, February 20, and February 26, 2013, supplements contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 27, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 22, 2011, as supplemented by letters dated March 30, September 10 and 28, 2012, and January 3, 2013.

Brief description of amendment: The amendment revised the curves in Technical Specification (TS) 3.4.9, “RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,” to replace the 28 Effective Full Power Years (EFPY) restriction in TS Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 and the minimum temperature in Surveillance Requirement (SR) 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7. The amendment would include a set of updated P/T curves for pressure test, core not critical, and core critical conditions for 32 EFPY based on a fluence evaluation performed using NRC-approved fluence methodology. The new curves would show a shift of minimum operating temperature which allows the bolt-up and minimum temperatures specified for SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 to be changed from 80 degrees Fahrenheit (°F) to 70 °F.

Date of issuance: February 22, 2013.

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

Amendment No.: 245.

Renewed Facility Operating License No. DPR-46: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: March 6, 2012 (77 FR 13372). The supplemental letters dated March 30, September 10 and 28, 2012, and January 3, 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 February 22, 2013.

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 20, 2012, as supplemented on December 7, 2012.

Brief description of amendment: The amendment revises the MNGP Technical Specifications (TS) Section 1.0, “Definitions,” Section 3.4.9, “RCS [Reactor Coolant System] Pressure and Temperature (P-T) Limits,” and Section 5.6, “Administrative Controls.” The amendment revises the P-T limits based on a methodology documented in the SIR-05-044-A report, “Pressure-Temperature Limits Report [PTLR] Methodology for Boiling Water Reactors,” and relocates the revised P-T limits from the TS to the MNGP PTLR.

Date of issuance: February 27, 2013.

Effective date: This license amendment is effective as of the date of its date of issuance and shall be implemented within 180 days after start-up from the 2013 Refueling Outage.

Amendment No.: 172.

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

Date of initial notice in Federal Register: April 17, 2012 (77 FR 22815). The licensee's December 7, 2012, supplemental letter did not change the scope of the original amendment request, did not change the NRC staff's initial proposed finding of no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 27, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: February 10, 2012, as supplemented by letters dated October 1, 2012, and January 22, 2013.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) to establish the limiting condition for operation (LCO) requirements for the reactor protective system (RPS) actuation circuits in TS 2.15, “Instrumentation and Control Systems.” Specifically, the TS changes renumbered LCOs 2.15(1) through 2.15(4) to 2.15.1(1) through 2.15.1(4), renumbered LCO 2.15(5) to LCO 2.15.3 with an associated Table 2-6, “Alternate Shutdown and Auxiliary Feedwater Panel Functions,” and implemented a new LCO 2.15.2 for the RPS logic and trip initiation channels. The amendment also revised the TS Table of Contents to reflect the renumbering and addition of the LCO for the RPS logic and trip initiation channels and the new Table 2-6.

Date of issuance: February 28, 2013.

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

Amendment No.: 270.

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

Date of initial notice in Federal Register: August 7, 2012 (77 FR 47128). The supplemental letters dated October 1, 2012, and January 22, 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 February 28, 2013.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of application for amendment: March 1, 2012, as supplemented by letter dated December 21, 2012.

Brief description of amendments: The proposed amendment would make miscellaneous changes to the Technical Specifications (TS) and Facility Operating License (FOL) including: (1) Correction of typographical errors; (2) deletion of historical requirements that have expired; (3) corrections of errors or omissions from previous license amendment requests; and (4) updating of components lists to reflect current plant design.

Date of issuance: February 25, 2013.

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

Amendment No.: 193.

Renewed Facility Operating License No. NPF-57: The amendment revised the TSs and the Facility Operating License.

Date of initial notice in Federal Register: April 3, 2012 (77 FR 20075). The letter dated December 21, 2012, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application.

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

No significant hazards consideration comments received: No.

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

Date of amendment request: January 15, 2013.

Brief description of amendment: The amendment authorizes a departure from the Virgil C. Summer Nuclear Station Units 2 and 3 plant-specific Design Control Document (DCD) Tier 2* material incorporated into the Updated Final Safety Analysis Report (UFSAR) to revise the requirements for shear reinforcement spacing in the nuclear island basemat below the auxiliary building.

Date of issuance: February 26, 2013.

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

Amendment No.: Unit 2—1, and Unit 3—1.

Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: January 25, 2013 (78 FR 5511).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 26, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 18, 2013.

Brief description of amendment: The amendment authorizes a departure from the VCSNS Units 2 and 3 plant-specific Design Control Document (DCD) Tier 2* material incorporated into the Updated Final Safety Analysis Report (UFSAR) by revising the structural criteria code for anchoring of headed shear reinforcement bar within the nuclear island basemat.

Date of issuance: March 1, 2013.

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

Amendment No.: Unit 2—2, and Unit 3—2.

Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: January 29, 2013 (78 FR 6145).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 1, 2013.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 15, 2013.

Brief description of amendment: The proposed amendment would depart from VEGP Units 3 and 4 plant-specific Design Control Document (DCD) Tier 2* material incorporated into the Updated Final Safety Analysis Report (UFSAR) to clarify the requirements for shear reinforcement spacing in the nuclear island basemat below the auxiliary building. The proposed change would modify the provisions for maximum spacing of the shear reinforcement in the basemat below the auxiliary building.

Date of issuance: February 26, 2013.

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

Amendment No.: Unit 3—4, and Unit 4—4.

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

Date of initial notice in Federal Register: January 25, 2013 (78 FR 5508).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 26, 2013.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 11th day of March 2013.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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