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Notice

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

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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 that 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 June 27, 2013 to July 10, 2013. The last biweekly notice was published on July 9, 2013 (78 FR 41118).

ADDRESSES:

You may submit comment by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2103-0158. Address questions about NRC dockets to Carol Gallagher; telephone: 301-492-3668; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: 3WFN-06A-44MP, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

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SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2013-0158 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and is publicly available, by the following methods:

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0158.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. 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-2013-0158 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.Start Printed Page 44168

The NRC cautions you not to include identifying or contact information in comment submissions that you do not want to be publicly disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS, and the NRC does not 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 in their comment submissions that they do not want to be publicly disclosed. Your request should state that the NRC will not 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 Title 10 of the Code of Federal Regulations (10CFR), Section 50.92, 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. NRC regulations are accessible electronically from the NRC Library on the NRC 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 Start Printed Page 44169governmental 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 hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1) a digital information (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 Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC 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 Web site at http://www.nrc.gov/​site-help/​e-submittals.html, by email at MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the 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 Start Printed Page 44170NRC'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 PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov.

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

Date of amendment request: May 23, 2013.

Description of amendment request: The proposed change would modify Technical Specifications (TS) to risk-inform requirements regarding selected Required Action End States. Specifically, the proposed change would permit an end state of Mode 4 rather than an end state of Mode 5 contained in the current TS. The proposed changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Technical Change Traveler 432-A Revision 1, “Change in Technical Specifications End States WCAP-16294.” This traveler revised the Improved Standard Technical Specifications.

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

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

Response: No.

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

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

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

Response: No.

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

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

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

Response: No.

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

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

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

Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Robert Beall, Acting.

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

Date of amendment request: March 26, 2013.

Description of amendment request: The amendment request would incorporate the NRC-approved Technical Specifications Task Force (TSTF) change traveler TSTF-431, Revision 3, “Change in Technical Specifications End States (BAW-2441),” and modify the Technical Specification (TS) requirements for end states associated with the implementation of the approved B&W Owners Group (B&WOG) Topical Report BAW-2441-A, Revision 2, “Risk-Informed Justification for LCO End-State Changes,” January 2004, as well as Required Actions revised by a specific Note in TSTF-431, Revision 3. The TS Actions End States modifications would permit, for some systems, entry into a hot shutdown (Mode 4) end state rather than a cold shutdown (Mode 5) end state that is the current TS requirement.

The NRC issued a “Notice of Availability of the Models for Plant-Specific Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-431, Revision 3, ‘Change in Technical Specifications End States (BAW-2441),’ ” in the Federal Register on December 6, 2010 (75 FR 75705-75706), which included the no significant hazards consideration, safety evaluation, and required commitments for the proposed changes as part of the consolidated line item improvement process (CLIIP).

In its application dated March 26, 2013, the licensee has concluded that the technical basis presented in the TSTF proposal and the safety evaluation are applicable to Arkansas Nuclear One, Unit 1, and the proposed amendment is consistent with the Standard Technical Specifications (STS) changes described in TSTF-431, Revision 3, but with certain variations and/or deviations from TSTF-431, Revision 3.

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

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

Response: No.

The proposed change allows a change to certain required end states when the Technical Specification (TS) Completion Times (CTs) for remaining in power operation are exceeded. Most of the requested TS changes are to permit an end state of hot shutdown (Mode 4) rather than an end state of cold shutdown (Mode 5) contained in the current TS. The request was limited to: 1) those end states where entry into the shutdown mode is for a short Start Printed Page 44171interval, 2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS, and 3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical. Risk insights from both the qualitative and quantitative risk assessments were used in specific TS assessments. Such assessments are documented in Sections 4 and 5 of BAW-2441-A, Revision 2, “Risk Informed Justification for LCO end-state Changes,” for B&W Plants. The assessments provide an integrated discussion of deterministic and probabilistic issues, focusing on specific TSs, which are used to support the proposed TS end state and associated restrictions. The staff finds that the risk insights support the conclusions of the specific TS assessments. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting proposed TSTF-431, Revision 3, are no different than the consequences of an accident prior to its adoption. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). If risk is assessed and managed, allowing a change to certain required end states when the TS Completion Times for remaining in power operation are exceeded; i.e., entry into hot shutdown rather than cold shutdown to repair equipment, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change and the commitment by the licensee to adhere to the guidance in TSTF-IG-07-01, Implementation Guidance for TSTF-431, Revision 1, “Changes in Technical Specifications end states, BAW-2441-A,” will further minimize possible concerns.

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

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

Response: No.

The proposed change allows, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed. The B&WOG's risk assessment approach is comprehensive and follows staff guidance as documented in [NRC Regulatory Guide (RG) 1.174, Revision 1, “An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis,” November 2002, and RG 1.177, “An Approach For Plant-Specific, Risk-Informed Decision Making: Technical Specifications,” August 1998]. In addition, the analyses show that the criteria of the three-tiered approach for allowing TS changes are met. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A risk assessment was performed to justify the proposed TS changes. The net change to the margin of safety is insignificant.

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

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

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

NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2 (ANO-2), Pope County, Arkansas

Date of amendment request: December 17, 2012.

Description of amendment request: The licensee has requested NRC review and approval for adoption of a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in NRC Regulatory Guide (RG) 1.205, Revision 1, “Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,” December 2009. The license amendment request follows Nuclear Energy Institute (NEI) 04-02, Revision 2, “Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c),” April 2008. This submittal describes the methodology used to demonstrate compliance with, and transition to, National Fire Protection Association (NFPA) 805, and includes regulatory evaluations, probabilistic risk assessment, change evaluations, proposed modifications for non-compliances, and supporting attachments.

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

Criterion 1

The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated. Operation of Arkansas Nuclear One, Unit 2 (ANO-2) in accordance with the proposed amendment does not result in a significant increase in the probability or consequences of accidents previously evaluated. The proposed amendment does not affect accident initiators or precursors as described in the ANO-2 Safety Analysis Report (SAR), nor does it adversely alter design assumptions, conditions, or configurations of the facility, and it does not adversely impact the ability of structures, systems, or components (SSCs) to perform their intended function to mitigate the consequences of accidents described and evaluated in the SAR. The proposed changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions as required by the accident analysis. The SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition will remain capable of performing their design functions.

The purpose of this amendment is to permit ANO-2 to adopt a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance contained in Regulatory Guide (RG) 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection requirements that are an acceptable alternative to the 10 CFR Part 50, Appendix R, fire protection features (69 FR 33536; June 16, 2004).

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

NFPA 805, taken as a whole, provides an acceptable alternative for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 10 CFR Part 50, meets the underlying intent of the NRC's existing fire protection regulations and guidance, and achieves defense-in-depth along with the goals, performance objectives, and performance criteria specified in NFPA 805, Chapter 1. In addition, if there are any increases in core damage frequency (CDF) or risk as a result of the transition to NFPA 805, the increase will be small, bounded by the delta risk Start Printed Page 44172requirements of NFPA 805, and consistent with the intent of the Commission's Safety Goal Policy.

Engineering analyses, which may include engineering evaluations, probabilistic risk assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been met. The SAR documents the analyses of design basis accidents (DBAs) at ANO-2. All accident analysis acceptance criteria will continue to be met with the proposed amendment. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions for the radiological consequence evaluations in the ANO-2 SAR. In addition, the applicable radiological dose acceptance criteria will continue to be met.

Based on the above, the implementation of this amendment to transition the Fire Protection Plan (FPP) at ANO-2 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a significant increase in the probability of any accident previously evaluated. In addition, all equipment required to mitigate an accident remains capable of performing the assumed function. Therefore, the consequences of any accident previously evaluated are not significantly increased with the implementation of this amendment.

Criterion 2

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

Operation of ANO-2 in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Previously analyzed accidents with potential offsite dose consequences were included in the evaluation of the transition to NFPA 805. The proposed amendment does not impact these accident analyses. The proposed change does not alter the requirements or functions for systems required during accident conditions as assumed in the licensing basis analyses and/or DBA [design-basis accident] radiological consequences evaluations.

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

The requirements in NFPA 805 address only fire protection and the impacts of fire effects on the plant have been evaluated. The proposed fire protection program changes do not involve new failure mechanisms or malfunctions that could initiate a new or different kind of accident beyond those already analyzed in the SAR. Based on this, as well as the discussion above, the implementation of this amendment to transition the FPP at ANO-2 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3

The Proposed Change Does Not Involve a Significant Reduction in a Margin of safety.

Operation of ANO-2 in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. The transition to a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed in the SAR to mitigate accidents. The proposed change does not adversely impact systems that respond to safely shut down the plant and maintain the plant in a safe shutdown condition. In addition, the proposed amendment will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without implementation of appropriate compensatory measures.

The risk evaluations for plant changes, in part as they relate to the potential for reducing a safety margin, were measured quantitatively for acceptability using the delta risk (i.e., ΔCDF and ΔLERF) criteria from Section 5.3.5, “Acceptance Criteria,” of NEI 04-02, as well as the guidance contained in RG 1.205. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based methods of NFPA 805 do not result in a significant reduction in the margin of safety. As such, the proposed changes are evaluated to ensure that risk and safety margins are kept within acceptable limits. Based on the above, the implementation of this amendment to transition the FPP at ANO-2 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), will not significantly reduce a margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas

Date of amendment request: March 26, 2013.

Description of amendment request: The amendment would incorporate the NRC-approved Technical Specifications Task Force (TSTF) change traveler TSTF-422, Revision 2, “Change in Technical Specifications End States (CE NPSD-1186).” The proposed amendment would modify Technical Specifications (TS) to risk-inform requirements regarding selected Required Action End States.

The NRC issued a “Notice of Availability (NOA) of the Models For Plant-Specific Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-422, Revision 2, `Change In Technical Specifications End States (CE NPSD-1186),' For Combustion Engineering (CE) Pressurized Water Reactor (PWR) Plants Using the Consolidated Line Item Improvement Process (CLIIP),” in the Federal Register on April 7, 2011 (76 FR 19510), which included the no significant hazards consideration, safety evaluation, and required commitments for the proposed changes as part of the consolidated line item improvement process (CLIIP).

In its application dated March 26, 2013, the licensee has concluded that the technical basis presented in the TSTF proposal and the safety evaluation are applicable to Arkansas Nuclear One, Unit 2, and the proposed amendment is consistent with the Standard Technical Specifications (STS) changes described in TSTF-422, Revision 2, but with certain variations and/or deviations from TSTF-422, Revision 2.

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

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

Response: No.

The proposed change allows a change to certain required end states when the Technical Specification (TS) Completion Times (CTs) for remaining in power operation are exceeded. Most of the requested TS changes are to permit an end state of hot shutdown (Mode 4) rather than an end state of cold shutdown (Mode 5) Start Printed Page 44173contained in the current TS. The request was limited to: (1) those end states where entry into the shutdown mode is for a short interval; (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS; and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical. Risk insights from both the qualitative and quantitative risk assessments were used in specific TS assessments. Such assessments are documented in Section 5.5 of CE NPSD-1186, Rev 0, “Technical Justification for the Risk-Informed Modification to Selected Required Action End States for CEOG [Combustion Engineering Owners Group] Member PWRs.” The assessments provide an integrated discussion of deterministic and probabilistic issues, focusing on specific TSs, which are used to support the proposed TS end state and associated restrictions. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting proposed TSTF-422 are no different than the consequences of an accident prior to adopting TSTF-422. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing a change to certain required end states when the TS CTs for remaining in power operation are exceeded, i.e., entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change and the commitment by the licensee to adhere to the guidance in WCAP-16364-NP, Revision 2, “Implementation Guidance for Risk Informed Modification to Selected Required Action End States at Combustion Engineering NSSS Plants (TSTF-422),” will further minimize possible concerns.

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

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

Response: No.

The proposed change allows, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed. The CEOG's risk assessment approach is comprehensive and follows NRC staff guidance as documented in [NRC Regulatory Guide (RG) 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision Making on Plant Specific Changes to the Licensing Basis,” August 1998, and RG 1.177, “An Approach for Pant Specific Risk-Informed Decision Making: Technical Specifications,” August 1998.]. In addition, the analyses show that the criteria of the three-tiered approach for allowing TS changes are met. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A risk assessment was performed to justify the proposed TS changes. The net change to the margin of safety is insignificant.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: May 21, 2013.

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

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

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 NSHC, which is presented below:

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

The proposed changes for St. Lucie Units 1 and 2 revise the MTC Technical Specification 4.1.1.4.1 and 4.1.1.4.2 for each Unit, to implement the requirements of the topical report WCAP-16011-P-A, STAR Program.

The MTC is not an initiator to any accident previously evaluated. Therefore, there is no significant increase in the probability of any accident previously evaluated. The MTC is an input to the accident analyses used to predict plant behavior in the event of an accident. The MTC limits specified in the Technical Specifications/COLR [core operating limit report] remain unchanged. WCAP-16011-P-A demonstrated, and the NRC concurred, that the modified MTC verification is adequate to ensure that MTC stays within the limits. The consequences of an accident after adopting TSTF-486 are no different than the consequences of an accident prior to adoption. Likewise, the deviations from the implementation of TSTF-486 requirements being adopted in this license amendment do not have any effect on the probability of occurrence or consequences of accidents previously evaluated.

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

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

No new or different accidents will result from implementation of the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different operating requirements or eliminate any existing requirements. The changes do not alter limits and assumptions made in the safety analysis. The proposed Start Printed Page 44174changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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?

TSTF-486 provides the means and requirements for CE-designed plants to implement the previously approved WCAP-16011-P-A for MTC verification at startup. MTC is a parameter controlled in the licensee's TS/COLR, including surveillance requirements. As stated previously, WCAP-16011-P-A describes methods to reduce the requirements for startup testing. The proposed changes to the TS, supported by TSTF-486, have been reviewed and found to be consistent with WCAP-16011-P-A. The changes in the license amendment which deviate from TSTF-486 requirements are justified to be acceptable and do not affect the margin of safety. The MTC limits are unaffected and an acceptable method will be used to verify the MTC to be within its limit. 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Jessie F. Quichocho.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: April 25, 2013.

Description of amendment request: The proposed license amendment request would revise certain requirements from Section 5, “Administrative Controls,” of the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITSs). The revisions would include the following sections: 5.1 “Responsibility;” 5.2 “Organization;” 5.6 “Procedures, Programs and Manuals;” 5.7 “Reporting Requirements;” and 5.8 “High Radiation Area,” which are no longer applicable, as CR-3 is in a permanently defueled condition.

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 for each proposed change, which is presented below:

A. ITS Section 5.1.1:

This section defines the responsible position for overall unit operation and for approval of each proposed test, experiment, or modification to systems or equipment that affect stored nuclear fuel and fuel handling. The responsible position title is changed from the Plant General Manager to the Plant Manager.

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

No. The change reflects that the remaining credible accident is a fuel handling accident or loss of spent fuel cooling. The change in the position title of the responsible person is administrative and cannot increase the probability or consequences of a fuel handling accident.

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

No. This changes reflects an organizational change to transition from an operating plant to a permanently defueled plant. Such an administrative change cannot create a new or different kind of accident.

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

No. The position title proposed here does not involve any physical plant limits or parameters and therefore cannot affect any margin of safety.

B. ITS Section 5.1.2:

This section identifies the responsibilities for the control room command function associated with Modes of plant operation, and is based on personnel positions and qualifications for an operating plant. It identifies the need for a delegation of authority for command in an operating plant when the principal assignee leaves the control room.

This section is being changed to eliminate the MODE dependency for this function and personnel qualifications associated with an operating plant. The proposed change establishes the Shift Supervisor as having command of the shift.

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

No. This is a change to the requirements for control room staffing. In a permanently defueled plant, the fuel handling building accident is the only credible accident previously evaluated. This action cannot increase the probability or consequences of a fuel handling accident.

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

No. The changes proposed here for control room staffing cannot create a new or different kind of accident since they do not change the function of any plant structures, systems, or components.

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

No. The changes proposed here for control room staffing do not directly involve any limits or parameters and therefore cannot affect ant margin of safety.

C. ITS Section 5.2.1.a:

The introduction to this section identifies that organizational positions are established that are responsible for the safety of the nuclear plant.

This is changed to require that positions be established that are responsible for the safe storage and handling of nuclear fuel. This change removes the implication that CR-3 can return to operation.

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

No. This change in the description of functional responsibility of organizational positions places emphasis on the safe storage and handling of nuclear fuel. This focus on their principal responsibility cannot increase the probability or consequences of a fuel handling accident.

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

No. This change in the description of functional responsibility of organizational positions cannot create a new or different kind of accident since they do not change the function of any plant structures, systems, or components.

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

No. This change does not directly involve any physical limits or parameters and therefore cannot affect any margin of safety.

D. ITS Section 5.1.2.b:

This section identifies the organizational position responsible for overall nuclear plant safety, for the safe operation of the plant, and for control of activities necessary for the safe operation and maintenance of the plant.

This section is being changed to recognize that the safety concerns for a permanently defueled plant are for the safe storage and handling of nuclear fuel. It changes responsibility for overall safety for storage and handling of nuclear fuel to the Decommissioning Director. It changes responsibility for control over onsite activities necessary for safe handling and storage of nuclear fuel to the Plant Manager.

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

No. This change in the description of functional responsibility of organizational positions places emphasis on the safe storage and handling of nuclear fuel. This focus on their principal responsibility cannot increase the probability or consequences of a fuel handling accident.

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

No. This change in the description of functional responsibility of organizational positions cannot create a new or different Start Printed Page 44175kind of accident since they do not change the function of any plant structures, systems, or components.

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

No. This change does not directly involve any physical limits or parameters and therefore cannot affect any margin of safety.

E. ITS Section 5.2.1.c:

This paragraph addresses the requirement for organizational independence of the operations, health physics, and quality assurance personnel from operating pressures.

This is changed to replace “operating staff” with “Certified Fuel Handlers,” and to replace “their independence from operating pressures” to “their ability to perform their assigned functions.”

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

No. This change continues to ensure that personnel in specifically identified positions retain independence from organizational pressures and will not increase the probability or occurrence of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components there it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

F. ITS Section 5.2.2.a:

This paragraph addresses that one auxiliary nuclear operator must be assigned to the operating shift whenever fuel is in the reactor.

Since this can never occur again at CR-3, the minimum requirement is changed to a minimum crew compliment of one Shift Supervisor and one Non-certified Operator.

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

No. This change, in conjunction with new paragraph 5.2.2.e, continues to ensure that personnel trained and qualified for the safe handling and storage of nuclear fuel are onsite. This cannot increase the probability or consequences of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

G. ITS Section 5.2.2.b:

This paragraph addresses the conditions under which the minimum shift compliment may be reduced. It contains a reference to 10 CFR 50.54(m) which establishes the minimum requirements for a licensed operating staff for facility operation.

This reference is removed since CR-3 will not return to operation in the future, and the requirement for licensed operating personnel will no longer be required to protect public health and safety.

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

No. This change continues to ensure that the minimum shift compliment of qualified personnel will not be decreased for more than a limited period. It removes the qualification requirements for personnel who are capable of responding to operating plant transients and accidents. This does not involve an increase in the probability or consequences of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

H. ITS Section 5.2.2.c:

This paragraph establishes the requirement for one licensed Reactor Operator to be in the control room when fuel is in the reactor and for one Senior Reactor Operator to be in the control room during operating Modes 1-4.

The change establishes the requirements for either a Non-certified operator or Certified Fuel handler to be in the control room when fuel is stored in the pools.

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

No. This change continues to ensure that personnel trained and qualified for the handling and storage of nuclear fuel man the control room. This cannot increase the probability or consequences of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

I. ITS Section 5.2.2.d:

This paragraph established the requirement for a person qualified in Radiation Protection procedures to be onsite when fuel is in the reactor.

This paragraph is deleted, since CR-3 is no longer authorized to have fuel in the reactor.

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

No. This administrative change cannot affect the probability of a fuel handling accident. The consequences of a fuel handling accident are governed by the characteristics of the fuel element and are not affected by the presence or absence of radiation protection trained personnel.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

J. ITS Section 5.2.2.d (New):

A new paragraph is added to establish the requirement for having oversight of fuel handling operations to be performed by a Certified Fuel Handler.

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

No. Certified Fuel Handlers are specifically trained and qualified to safely handle irradiated fuel. Applying these qualifications to fuel movement ensures that the probability or consequences of a fuel handling accident are not increased.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

K. ITS Section 5.2.2.e (New):

A new paragraph is added to establish that the Shift Supervisor must be a Certified Fuel Handler.

In the permanently defueled plant, the Certified Fuel Handler is the senior position on the operating crew. It is not necessary for the Shift Supervisor to hold a Senior Reactor Operator license if the plant cannot operate to generate power.

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

No. Certified Fuel Handlers are specifically trained and qualified to safely handle irradiated fuel. Applying these qualifications to the supervision of fuel movement ensures Start Printed Page 44176that the probability or consequences of a fuel handling accident are not increased.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

L. ITS Section 5.3.1:

This paragraph is changed to remove the requirements for the Shift Technical Advisor since that position is only required for a plant authorized for power operations.

The paragraph retains the previous requirements for the personnel filling unit staff positions meet or exceed the minimum qualifications of ANSI [American National Standard Institute] N18.1, 1971, and the Radiation Protection Manager meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

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

No. The Shift Technical Advisor position was established to assist the control room operating personnel to diagnose the cause and advise on the response to operating transients and accidents. The absence of a staff member with those qualifications does not change the probability or consequences of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any physical equipment limits or parameters and therefore cannot affect any margin of safety.

M. ITS Section 5.3.2:

This new paragraph is added to identify that responsibility for the training and retraining of Certified Fuel Handlers is assigned to the Plant Manager.

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

No. This section recognizes the importance of establishing and maintaining Certified Fuel Handler qualifications and assigns a manager responsibility for this program. Training and retraining Certified Fuel Handlers specifically trained to safely handle nuclear fuel will not increase the probability or consequences of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any physical limits or parameters and therefore cannot affect any margin of safety.

N. ITS Section 5.6.1.1.a:

This section states the requirement for procedures to be established, implemented and maintained covering various plant activities.

The scope is reduced to procedures applicable to the safe handling and storage of nuclear fuel.

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

No. The procedures necessary for the safe handling of nuclear fuel are included in the group of procedures applicable to the safe storage of nuclear fuel. With these procedures in effect for fuel handling, the probability or consequences of a fuel handling accident will not be increased.

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

No. The applicable procedures for the safe storage of nuclear fuel will direct the correct use of fuel handling equipment. These procedures are currently in place and have been used effectively for the safe handling of fuel. These procedures will not direct the use of plant structures, systems, or components in a different manner, therefore, they cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

O. ITS Section 5.6.2.3:

In this section, the authority for approval of changes to the Offsite Dose Calculation Manual (ODCM) is changed from the Plant General Manager to the Plant Manager consistent with the position title change in 5.1.1.

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

No. This is a change to the requirements for the position responsible for approving ODCM changes. In a permanently defueled plant, the fuel handling accident is the only credible accident previously evaluated. This action cannot increase the probability or consequences of a fuel handling accident.

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

No. The change proposed here, identifying a different position responsible for ODCM change approval, cannot create a new or different kind of accident since this does not change the function of any plant structures, systems, or components.

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

No. The changes proposed here for ODCM approval do not directly involve any limits or parameters for operating systems and therefore cannot affect any margin of safety.

P. ITS Section 5.6.2.4: Primary Coolant Sources Outside Containment

This program was established to minimize leakage from portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident.

The program is being eliminated.

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

No. The fuel handling accident is the only credible accident for a permanently defueled plant. This change eliminates an inspection program that is no longer necessary to limit the consequences of operating transients and accidents. This change cannot increase the probability or consequences of the fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

Q. ITS Section 5.6.2.5: Component Cyclic or Transient Limit

This program provided controls to track cyclic and transient occurrences to ensure that components were maintained within their design limits.

This program is being eliminated.

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

No. Eliminating an administrative event tracking program cannot increase the probability of a fuel handling accident.

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

No. Eliminating an administrative event tracking program cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

R. ITS Section 5.6.2.8: Inservice Inspection Program

This program required periodic inspections, examinations, and tests of plant pressure boundary components to ensure their continued integrity for power operation.

This program is being eliminated.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?Start Printed Page 44177

No. The Inservice Inspection Program does not apply to nuclear fuel or fuel handling equipment. Therefore eliminating this program cannot increase the probability or occurrence of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. For an operating plant the Inservice Inspection Program provided confidence that plant systems that were either a potential source of an accident or transient or served to mitigate events continued to meet their physical requirements. For a permanently shutdown plant, no transient, or accident can occur, so ending this inspection program cannot affect any margin of safety.

S. ITS Section 5.6.2.10: Steam Generator (OTSG) Program

The Steam Generator Program established and implemented practices to ensure that OTSG tube integrity was maintained.

This program is being eliminated.

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

No. The condition of the steam generator tubes inside the containment has no effect on fuel handling in the auxiliary building within the spent fuel pools. Therefore, eliminating the program cannot increase the probability or occurrence of a fuel handling accident.

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

No. The CR-3 steam generators will remain out of service until removed from the plant. In this state, the condition of the steam generator tubes is immaterial and cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

T. ITS Section 5.6.2.11: Secondary Water Chemistry Program

This program provided controls for monitoring secondary water chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking.

This program is being eliminated.

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

No. The secondary piping systems do not interconnect with the fuel cooling or fuel handling systems. Therefore, eliminating the Secondary Water Chemistry Program cannot increase the probability or occurrence of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. The components this program was intended to protect will no longer function for power production. Therefore, eliminating this program cannot affect any margin of safety.

U. ITS Section 5.6.2.13: Explosive Gas and Storage Tank Radioactivity Monitoring Program

This program provided controls for potentially explosive gas mixtures contained in the Radioactive Waste Disposal (WD) System, and the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system.

This program is being eliminated.

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

No. This program is required for an operating plant where hydrogen and radioactive gases are created and must be controlled. Controlled release of any gases currently in the tanks, in accordance with existing procedures, will ensure there will be no hazard to public health and safety. Therefore, elimination of this program cannot increase the probability or consequences of a fuel handling accident.

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

No. This program is required for an operating plant where hydrogen and radioactive gases are created and must be controlled. Controlled release of any gases currently in the tanks, in accordance with existing procedures, will ensure there will be no hazard to public health and safety. Therefore, elimination of this program cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margins of safety.

V. ITS Section 5.6.2.18: Core Operating Limits Report (COLR)

This program established that core operating limits be established prior to each reload cycle.

This program is being eliminated.

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

No. This program for controlling the design and operation of the reactor core has no bearing on fuel storage after fuel has been moved into the spent fuel pools. Therefore, eliminating this program cannot increase the probability or occurrence of a fuel handling accident.

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

No. Since CR-3 can never load a core into the reactor again, eliminating this control program cannot create a new or different kind of accident.

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

No. Since CR-3 can never load a core into the reactor again, eliminating this control program cannot affect any margin of safety.

W. ITS 5.6.2.19: Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

This program ensured that RCS pressure and temperature limits, including heatup and cooldown rates, criticality, and hydrostatic and leak test limits, be established and documented in the PTLR.

This program is being eliminated.

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

No. This program contains no actions or limits that affect the storage or handling of nuclear fuel. Therefore, eliminating this program cannot increase the probability or occurrence of a fuel handling accident.

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

No. This report is no longer needed since the reactor coolant system is not subject to pressurization and the reactor contains no fuel. Therefore, eliminating this control program cannot create a new or different kind of accident.

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

No. The limits established in this report do not apply to nuclear fuel stored in the spent fuel pools. Therefore, eliminating this program cannot affect any margin of safety.

X. ITS Section 5.6.2.20: Containment Leakage Rate Testing Program

This program was established to implement the leakage rate testing of the containment.

This program is being eliminated in accordance with Regulatory Guide 1.1.84.

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

No. Since fuel can never be returned to the CR-3 containment, ending containment leakage rate testing cannot increase the probability or occurrence of a fuel handling accident.

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

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

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

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.Start Printed Page 44178

Y. ITS Section 5.7.2: Special Reports

This section is being eliminated.

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

No. Eliminating reporting requirements for programs that are no longer required or conditions that cannot exist in a permanently defueled plant cannot increase the probability or occurrence of a fuel handling accident.

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

No. Eliminating reporting requirements that are no longer required cannot create a new or different kind of accident.

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

No. Eliminating reporting requirements that are no longer required cannot affect any margin of safety.

Z. ITS Section 5.8.2: High Radiation Area Controls

Changes one of the personnel responsible for locked high radiation area key control from the Control Room Supervisor to the Shift Supervisor.

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

No. This is a change to the requirements for the position title responsible for key control. In a permanently defueled plant, the fuel handling accident is the only credible accident previously evaluated. This action cannot increase the probability or consequences of a fuel handling accident.

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

No. The change proposed here, identifying a different position title responsible for key control, cannot create a new or different kind of accident since they do not change the function of any plant structures, systems, or components.

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

No. The changes proposed here for key control do not directly involve any limits or parameters and therefore cannot affect any margin of safety.

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

Attorney for licensee: Kathryn B. Nolan, 550 South Tryon Street, Charlotte, North Carolina, 28202.

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 pdr.resource@nrc.gov.

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

Date of application for amendments: July 25, 2012.

Brief description of amendments: The amendments revise Technical Specifications (TSs) 3.4.19—“Steam Generator (SG) Tube Integrity,” 5.5.8—“Steam Generator (SG) Program,” and 5.6.7—“Steam Generator Tube Inspection Report” to apply the appropriate program attributes to the Unit 2 replacement steam generators that are planned for installation in fall 2013. The amendments also revise the PINGP Units 1 and 2 TSs to adopt the program improvements in Technical Specifications Task Force Traveler (TSTF) 510, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.”

Date of issuance: July 2, 2013.

Effective date: As of the date of issuance and shall be implemented within 60 days after reactor startup following Unit 2 steam generator replacements.

Amendment Nos.: 208 and 195.

Renewed Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 14, 2012 (77 FR 56881).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 2, 2013.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 20, 2013.

Brief description of amendment: The amendment authorizes a departure from the Vogtle Electric Generating Plant Units 3 and 4 plant-specific Design Control Document (DCD) material incorporated into the Updated Final Safety Analysis Report (UFSAR) by revising the structural analysis requirements to provide alternative requirements for development of headed reinforcement bars (T-heads) within the nuclear island structures above the basemat elevation.

Date of issuance: May 22, 2013.

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

Amendment No.: Unit 3-9 and Unit 4-9.Start Printed Page 44179

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

Date of initial notice in Federal Register: April 16, 2013 (78 FR 22573).

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

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 15th day of July 2013.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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

End Signature End Supplemental Information

[FR Doc. 2013-17370 Filed 7-22-13; 8:45 am]

BILLING CODE 7590-01-P