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Notice

Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving Proposed No Significant Hazards Considerations and Containing Sensitive Unclassified Non-Safeguards Information and Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information

Document Details

Information about this document as published in the Federal Register.

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Start Preamble

AGENCY:

Nuclear Regulatory Commission.

ACTION:

License amendment request; opportunity to comment, request a hearing, and petition for leave to intervene; order.

SUMMARY:

The U.S. Nuclear Regulatory Commission (NRC) received and is considering approval of six amendment requests. The amendment requests are for Fermi 2; Beaver Valley Power Station, Units 1 and 2; Monticello Nuclear Generating Plant; South Texas Project, Units 1 and 2; Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2; South Texas Project, Units 1 and 2; and Wolf Creek Generating Station. For each amendment request, the NRC proposes to determine that they involve no significant hazards consideration. In addition, each amendment request contains sensitive unclassified non-safeguards information (SUNSI).

DATES:

Comments must be filed by October 9, 2014. A request for a hearing must be filed by November 10, 2014. Any potential party as defined in § 2.4 of Title 10 of the Code of Federal Regulations (10 CFR), who believes access to SUNSI is necessary to respond to this notice must request document access by September 19, 2014.

ADDRESSES:

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

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0190. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

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FOR FURTHER INFORMATION CONTACT:

Mable Henderson, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-3760, email: Mable.Henderson@nrc.gov.

End Further Info End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2014-0190 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0190.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section.
  • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One Start Printed Page 53456White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2014-0190 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

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

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

II. Background

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

This notice includes notices of amendments containing SUNSI.

III. 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 10 CFR 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.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR Part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed within 60 days, 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 set forth 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 Start Printed Page 53457contention 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, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of any amendment.

B. Electronic Submissions (E-Filing)

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

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

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

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

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

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

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

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or Start Printed Page 53458home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii).

For further details with respect to this amendment action, see the application for amendment which is available for public inspection at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. 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.

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

Date of amendment request: July 2, 2014. A publicly-available version is in ADAMS under Accession No. ML14183B528.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The license amendment request pertains to the Cyber Security Plan (CSP) implementation schedule change in the completion date for Milestone 8. Milestone 8 pertains to the date that full implementation of the CSP for all safety, security, and emergency preparedness functions will be achieved.

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 amendment proposes a change to the Fermi 2 Cyber Security Plan (CSP) M8 full implementation date as set forth in the Fermi 2 CSP Implementation Schedule. The revision of the full implementation date for the Fermi 2 CSP does not involve modifications to any safety-related structures, systems or components (SSCs). The implementation schedule provides a timetable for fully implementing the Fermi 2 CSP. The CSP describes how the requirements of 10 CFR 73.54 are to be implemented to identify, evaluate, and mitigate cyber-attacks up to and including the design basis cyber-attack threat, thereby achieving high assurance that the facility's digital computer and communications systems and networks are protected from cyber-attacks. The revision of the Fermi 2 CSP Implementation Schedule will not alter previously evaluated design basis accident analysis assumptions, add any accident initiators, modify the function of the plant safety-related SSCs, or affect how any plant safety-related SSCs are operated, maintained, modified, tested, or inspected.

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?

Response: No.

The implementation of the Fermi 2 CSP does not introduce new equipment that could create a new or different kind of accident, and no new equipment failure modes are created. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed amendment.

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

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

Response: No.

The margin of safety is associated with the confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant pressure boundary, and containment structure) to limit the level of radiation to the public. The proposed amendment does not alter the way any safety-related SSC functions and does not alter the way the plant is operated.

The CSP provides assurance that safety-related SSCs are protected from cyber-attacks. The proposed amendment does not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. The proposed amendment has no effect on the structural integrity of the fuel cladding, reactor coolant pressure boundary, or containment structure. Based on the above considerations, the proposed amendment does not degrade the confidence in the ability of the fission product barriers to limit the level of radiation to the public.

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

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

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

NRC Branch Chief: David L. Pelton.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Units 1 and 2, (BVPS-1 and BVPS-2) Beaver County, Pennsylvania

Date of amendment request: December 23, 2013, as supplemented by letter dated February 14, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14002A086, and ML14051A499, respectively.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would change the BVPS-1 and BVPS-2 facility operating license and technical specifications. Specifically, the amendment requests review and approval to transition the fire protection licensing basis at BVPS, from Title 10 of the Code of Federal Regulations (10 CFR), Section 50.48(b), to 10 CFR 50.48(c), National Fire Protection Association (NFPA) 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants.” The adoption of NFPA-805, would provide BVPS with a risk-informed, performance-based fire protection program, and allow them to make changes to their fire protection program without prior NRC approval, only if the changes would not adversely affect the plant's ability to achieve and maintain safe shutdown in the event of a fire.

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

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

Response: No.Start Printed Page 53459

Operation of BVPS-1 and BVPS-2 in accordance with the proposed amendment does not increase the probability or consequences of accidents previously evaluated. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been satisfied. The Updated Final Safety Analysis Report (UFSAR) documents the analyses of design basis accidents (DBA) at BVPS-1 and BVPS-2. The proposed amendment does not adversely affect accident initiators nor alters design assumptions, conditions, or configurations of the facility and does not adversely affect the ability of structures, systems, or components (SSCs) to perform their design functions. 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 the proposed amendment is to permit BVPS-1 and BVPS-2 to adopt a new fire protection licensing basis, which complies with the requirements of 10 CFR 50.48(c) and the guidance in Regulatory Guide 1.205, Revision 1 [Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants (ADAMS Accession No. ML092730314)]. 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 50, Appendix R-required fire protection features (69 FR 33536; June 16, 2004).

Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been met. NFPA 805, taken as a whole, provides an acceptable alternative for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 10 CFR 50 and meets the underlying intent of the NRC's existing fire protection regulations and guidance. It also achieves defense in depth and the goals, performance objectives, and performance criteria specified in Chapter 1 of the standard and, if there are any increases in core damage frequency (CDF) or risk the increase will be small and consistent with the intent of the Commission's Safety Goal Policy.

Based on this, the implementation of the proposed amendment does not increase the probability of any accident previously evaluated. Equipment required to mitigate an accident remains capable of performing the assumed function. The proposed amendment will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The applicable radiological dose criteria will continue to be met. Therefore, the consequences of any accident previously evaluated are not increased with the implementation of the proposed amendment.

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

Response: No.

Operation of BVPS-1 and BVPS-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. Any scenario or previously analyzed accident with off-site dose was included in the evaluation of DBAs documented in the UFSAR. The proposed change does not alter the requirements or function for systems required during accident conditions. Implementation of the new fire protection licensing basis which complies with the requirements of 10 CFR 50.48(a) and 10 CFR 50.[4]8(c) and the guidance in Regulatory Guide 1.205, Revision 1 will not result in new or different accidents.

The proposed amendment does not adversely affect accident initiators or alter design assumptions, conditions, or configurations of the facility. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition remain capable of performing their design functions.

The purpose of the proposed amendment is to permit BVPS-1 and BVPS-2 to adopt a new fire protection licensing basis which complies with the requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance in Regulatory Guide 1.205, Revision 1. 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 50, Appendix R-required fire protection features (69 FR 33536; June 16, 2004). Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been met.

The requirements of NFPA 805 address only fire protection and the impacts of fire on the plant that have previously been evaluated. Based on this, the implementation of the proposed amendment does not create the possibility of a new or different kind of accident from any kind of accident previously evaluated. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety-related system as a result of this amendment. Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created with the implementation of this amendment.

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

Response: No.

Operation of BVPS-1 and BVPS-2 in accordance with the proposed amendment does not involve a significant reduction in the margin of safety. The risk evaluation of plant changes, as appropriate, were measured quantitatively for acceptability using the delta CDF and delta [Large Early Release Frequency] LERF criteria from Section 5.3.5 of NEI 04-02 and of Regulatory Guide 1.205, Revision 1. The proposed amendment does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed to mitigate accidents in the UFSAR. This amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition remain capable of performing their design functions.

The purpose of the proposed amendment is to permit BVPS-1 and BVPS-2 to adopt a new fire protection licensing basis, which complies with the requirements in 10 CFR 50.48(c) and the guidance in Regulatory Guide 1.205, Revision 1. 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 50, Appendix R-required fire protection features (69 FR 33536; June 16, 2004). Engineering analyses, which may include engineering evaluations, probabilistic safety assessments and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been met.

The proposed changes are evaluated to ensure that risk and safety margins are kept within acceptable limits. Therefore, the transition to NFPA 805 does not involve a significant reduction in the margin of safety. The requirements of NFPA 805 are structured to implement the NRC's mission of the protection of public health and safety, promote the common defense and security, and protect the environment. NFPA 805 is also consistent with the key principles for evaluating license basis changes as described in Regulatory Guide 1.174 [An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML100910006)] and is consistent with the defense in depth philosophy while maintaining sufficient safety margins.

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

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

Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

Acting NRC Branch Chief: Robert G. Schaaf.Start Printed Page 53460

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

Date of amendment request: July 15, 2013, as supplemented by letters dated January 31, 2014, March 12, 2014, April 29, 2014, and May 9, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML13200A185, ML14035A297, ML14077A291, ML14153A498, and ML14132A189, respectively.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would allow for a transition to the AREVA ATRIUM 10XM fuel design and implementation of AREVA safety analysis methods at the MNGP. The transition to the AREVA ATRIUM 10XM fuel design would permit use at Extended Power Uprate conditions with operation in the Maximum Extended Load Line Limit Analysis (MELLLA) power-flow operating domain. Specifically, NSPM proposed to revise Technical Specification (TS) 5.6.3, “Core Operating Limits Report (COLR),” to add AREVA analysis methodologies to the list of approved methods used in determining core operating limits. Northern States Power Company—Minnesota also proposes to (1) revise TS 2.1, “SL [Safety Limits],” to change the steam dome pressure associated with safety limits when using AREVA safety analysis methods, and (2) insert an editorial change to TS 4.2.1, “Reactor Core, Fuel Assemblies,” to add “water channel,” in addition to the current “water rod,” to reflect the design of the AREVA ATRIUM 10XM fuel assembly design feature.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below, with minor editorial revisions designated in brackets ([ ]):

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

Response: No.

Changing fuel designs and making an editorial change to TS will not increase the probability of a Loss of Coolant Accident (LOCA). The fuel cannot increase the probability of a primary coolant system breach or rupture, as there is not interaction between the fuel and the system piping. The fuel will continue to meet the 10 CFR 50.46, [“Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,”] limits. Therefore, the consequences of a LOCA will not be increased.

The probability of a Control Rod Drop Accident (CRDA) does not increase because the ATRIUM 10XM fuel channel is mechanically compatible with the co-resident fuel and existing control blade designs. The mechanical interaction and friction forces between the ATRIUM 10XM channel and control blades would not be higher than previous designs. In addition, routine plant testing includes confirmation of adequate control blade to control rod drive coupling. The probability of a CRDA is not increased with the use of ATRIUM 10XM fuel. CRDA consequences are evaluated on a cycle-specific basis, confirming the number of calculated rod failures remains within the Updated Safety Analysis Report (USAR) design basis. Similarly, changing the fuel design and making an editorial change to TS cannot increase the probability of an anticipated operation occurrence (AOO). As a passive component, the fuel does not interact with plant operating or control systems. Therefore, the fuel change cannot affect the initiators of the previously evaluated AOO transient events. Thermal limits for the new fuel will be determined on a cycle-specific basis, ensuring the specified acceptable fuel design limits continue to be met. Therefore, the consequences of a previously evaluated AOO will not increase.

A disposition of the plant's postulated accidents with radiological consequences indicated that the consequences of only two accidents could be affected by the proposed change in fuel type; the LOCA and the CRDA. Revised dose analyses using the approved Alternative Source Term methodology at Extended Power Uprate (EPU) rated power concluded that the change in fuel type resulted in small variations in the radiological source term for the reactor core and a corresponding slight difference in the overall dose consequences. At no location did the calculated dose increase more than two percent compared to previously-submitted radiological dose at EPU power levels. Dose consequences for the LOCA and CRDA with an ATRIUM 10XM fuel source term remained well below the regulatory limits of 10 CFR 50.67, [“Accident source term,”] and Regulatory Guide 1.183, [“Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.”]

The proposed change to Reactor Core Safety Limits involves a technical evaluation that demonstrates the range of applicability for the AREVA Critical Power Correlations will always bound the postulated pressures of plant transients using the AREVA safety analysis methodology. As a technical evaluation, this proposed change involves no physical change to a system, structure, component, or setpoint. Therefore, the proposed change in no way can affect the probability or the consequences of an accident previously evaluated.

[Therefore, the proposed activity 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 ATRIUM 10XM fuel product has been designed to maintain neutronic, thermal-hydraulic, and mechanical compatibility with the co-resident fuel designs currently in use at MNGP. The ATRIUM 10XM fuel has been designed to meet fuel licensing criteria specified in NUREG-0800, “Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants.” Compliance with these criteria ensures the fuel will not fail in an unexpected manner.

A change in fuel design and an editorial change to TS cannot create any new accident initiators because the fuel is a passive component having no direct influence on the performance of operating plant systems and equipment. Hence, a fuel design change cannot create a new type of malfunction leading to a new or different kind of accident. Consequently, the proposed fuel design change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Also, as a technical evaluation, the proposed change to Reactor Core Safety Limits involves no physical change to a system, structure, component, or setpoint. Therefore, this proposed change could in no way introduce a new physical interaction that would create the possibility of a new or different kind of accident from any accident previously evaluated.

[Therefore, the proposed change does not result in 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 ATRIUM 10XM fuel is designed to comply with the fuel licensing criteria specified in NUREG-0800. Cycle-specific and cycle-independent safety analyses are performed ensuring no fuel failures will occur as the result of anticipated operational occurrences, and dose consequences for accidents remain with the bounds of 10 CFR 50.67. Applicable regulatory margins and requirements are maintained.

The proposed change to Reactor Core Safety Limits is consistent with, and within the capabilities of the applicable NRC approved critical power correlation, and thus continues to ensure that valid critical power calculations are performed. No setpoints at which protective actions are initiated are altered by the proposed change. The proposed change does not alter the manner in which the safety limits are determined. This change is consistent with plant design and does not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.

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 Start Printed Page 53461proposes to determine that the requested amendments involve no significant hazards consideration.

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

NRC Branch Chief: David L. Pelton.

PSEG Nuclear LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2, Salem County, New Jersey

Date of amendment request: December 24, 2013, as supplemented by letter dated June 23, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14016A079 and ML14174B239, respectively.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would revise the Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2, Cyber Security Plan (CSP) Milestone 8 full implementation date as set forth in the Cyber Security Plan Implementation Schedule.

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

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

Response: No.

The amendment proposes a change to the PSEG Cyber Security Plan (CSP) Implementation Schedule Milestone 8 (M8) full implementation date as set forth in the PSEG CSP Implementation Schedule. The revision of the full implementation date for M8 does not involve modifications to any safety-related structures, systems, or components (SSCs). Rather, the implementation schedule provides a timetable for fully implementing the PSEG CSP. The CSP describes how the requirements of 10 CFR 73.54 are to be implemented to identify, evaluate, and mitigate cyber attacks up to and including the design basis cyber attack, thereby achieving high assurance that the facility's digital computer and communications systems and networks are protected from cyber attacks. The revision of the PSEG CSP Implementation Schedule will not alter previously evaluated design basis accident analysis assumptions, add any accident initiators, modify the function of plant safety-related SSCs, or affect how any plant safety-related SSCs are operated, maintained, modified, tested, or inspected.

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

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

Response: No.

The implementation of the PSEG CSP does not introduce new equipment that could create a new or different kind of accident, and no new equipment failure modes are created. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed amendment.

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

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

Response: No.

The margin of safety is associated with the confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant pressure boundary, and containment structure) to limit the level of radiation (dose or exposure) to the public. The proposed amendment does not alter the way any safety-related SSCs function and does not alter the way the plant is operated. The CSP provides assurance that safety-related SSCs are protected from cyber attacks. The proposed amendment does not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. The proposed amendment has no effect on the structural integrity of the fuel cladding, reactor coolant pressure boundary, or containment structure. Based on the above considerations, the proposed amendment does not degrade the confidence in the ability of the fission product barriers to limit the level of radiation (dose or exposure) to the public.

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

Acting NRC Branch Chief: Robert G. Schaaf.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: May 8, 2014, as supplemented by letter dated May 8, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14142A018 and ML14142A013, respectively.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendments would revise the South Texas Project (STP) Cyber Security Plan (CSP) Milestone 8 full implementation date as set forth in the STP CSP Implementation Schedule. The amendments would also revise paragraph 2.H of Facility Operating License Nos. NPF-76 for STP, Unit 1 and NFP-80 for STP, Unit 2, by incorporating the revised CSP implementation schedule.

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

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

Response: No.

The amendment proposes a change to the STP CSP Milestone 8 full implementation date as set forth in the STP CSP Implementation Schedule. The revision of the full implementation date for the STP CSP does not involve modifications to any safety-related structures, systems, or components (SSCs). Rather, the implementation schedule provides a timetable for fully implementing the STP CSP. The CSP describes how the requirements of 10 CFR 73.54 are to be implemented to identify, evaluate, and mitigate cyber attacks up to and including the design basis cyber attack threat, thereby achieving high assurance that the facility's digital computer and communications systems and networks are protected from cyber attacks. The revision of the STP CSP Implementation Schedule will not alter previously evaluated design basis accident analysis assumptions, add any accident initiators, modify the function of the plant safety-related SSCs, or affect how any plant safety related SSCs are operated, maintained, modified, tested, or inspected.

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

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

Response: No.

The implementation of the STP CSP does not introduce new equipment that could create a new or different kind of accident, and no new equipment failure modes are created. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed amendment.

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

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

Response: No.Start Printed Page 53462

The margin of safety is associated with the confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant pressure boundary, and containment structure) to limit the level of radiation to the public. The proposed amendment does not alter the way any safety related SSC functions and does not alter the way the plant is operated. The STP CSP provides assurance that safety-related SSCs are protected from cyber attacks. The proposed amendment does not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. The proposed amendment has no effect on the structural integrity of the fuel cladding, reactor coolant pressure boundary, or containment structure. Based on the above considerations, the proposed amendment does degrade the confidence in the ability of the fission product barriers to limit the level of radiation to the public.

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

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

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.

NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: February 26, 2014. A publicly-available version is in ADAMS under Accession No. ML14064A328.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendment would revise Technical Specification 5.6.5, “CORE OPERATING LIMITS REPORT (COLR),” to incorporate Westinghouse Electric Company LLC's topical report WCAP-16009-P-A, “Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),” January 2005, to the list of analytical methods used to determine the core operating limits. A non-proprietary version of the topical report, designated as WCAP-16009-NP-A, is available in ADAMS under Accession Nos. ML050910159 and ML050910161.

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

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

Response: No.

The proposed change revises Specification 5.6.5 to incorporate a new large break LOCA [loss-of-coolant accident] analysis methodology. Specifically, the proposed change adds WCAP-16009-P-A to Specification 5.6.5b. as a method used for establishing core operating limits.

Accident analyses are not accident initiators; therefore, the proposed change does not involve a significant increase in the probability of an accident. The analyses using ASTRUM demonstrated that the acceptance criteria in 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,” were met. Large break LOCA analyses performed consistent with the methodology in Nuclear Regulatory Commission (NRC) approved WCAP-16009-P-A, including applicable assumptions, limitations and conditions, demonstrate that 10 CFR 50.46 acceptance criteria are met; thus, this change does not involve a significant increase in the consequences of an accident. No physical changes to the plant are associated with the proposed change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change revises Specification 5.6.5 to incorporate a new large break LOCA analysis methodology. Specifically, the proposed change adds WCAP-16009-P-A to Specification 5.6.5b. as a method used for establishing core operating limits. There are no physical changes being made to the plant as a result of using the Westinghouse ASTRUM analysis methodology in WCAP-16009-P-A for performance of the large break LOCA analyses. Large break LOCA analyses performed consistent with the methodology in NRC approved WCAP-16009-P-A, including applicable assumptions, limitations and conditions; demonstrate that 10 CFR 50.46 acceptance criteria are met. No new modes of plant operation are being introduced. The configuration, operation, and accident response of the structures or components are unchanged by use of the new analysis methodology. Analyses of transient events have confirmed that no transient event results in a new sequence of events that could lead to a new accident scenario. The parameters assumed in the analyses are within the design limits of existing plant equipment.

In addition, employing the Westinghouse ASTRUM large break LOCA analysis methodology does not create any new failure modes that could lead to a different kind of accident. The design of systems remains unchanged and no new equipment or systems have been installed which could potentially introduce new failure modes or accident sequences. No changes have been made to instrumentation actuation setpoints. Adding the reference to WCAP-16009-P-A in Specification 5.6.5b. is an administrative change that does not create the possibility of a new or different kind of accident.

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

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

Response: No.

The proposed change revises Specification 5.6.5 to incorporate a new large break LOCA analysis methodology. Specifically, the proposed change adds WCAP-16009-P-A to Specification 5.6.5b. as a method used for establishing core operating limits. The analyses using ASTRUM demonstrated that the applicable acceptance criteria in 10 CFR 50.46 are met. Margins of safety for large break LOCAs include quantitative limits for fuel performance established in 10 CFR 50.46. These acceptance criteria are not being changed by this proposed new methodology. Large break LOCA analyses performed consistent with the methodology in NRC approved WCAP-16009-P-A, including applicable assumptions, limitations and conditions, demonstrate that 10 CFR 50.46 acceptance criteria are met; thus, this change does not involve a significant reduction in a margin of safety.

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

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

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.Start Printed Page 53463

Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information for Contention Preparation

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

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Units 1 and 2, (BVPS-1 and BVPS-2) Beaver County, Pennsylvania

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

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

PSEG Nuclear LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2, Salem County, New Jersey

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information for Contention Preparation

A. This Order contains instructions regarding how potential parties to this proceeding may request access to documents containing SUNSI.

B. Within 10 days after publication of this notice of hearing and opportunity to petition for leave to intervene, any potential party who believes access to SUNSI is necessary to respond to this notice may request such access. A “potential party” is any person who intends to participate as a party by demonstrating standing and filing an admissible contention under 10 CFR 2.309. Requests for access to SUNSI submitted later than 10 days after publication of this notice will not be considered absent a showing of good cause for the late filing, addressing why the request could not have been filed earlier.

C. The requester shall submit a letter requesting permission to access SUNSI to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, and provide a copy to the Associate General Counsel for Hearings, Enforcement and Administration, Office of the General Counsel, Washington, DC 20555-0001. The expedited delivery or courier mail address for both offices is: U.S. Nuclear Regulatory Commission, 11555 Rockville Pike, Rockville, Maryland 20852. The email address for the Office of the Secretary and the Office of the General Counsel are Hearing.Docket@nrc.gov and OGCmailcenter@nrc.gov, respectively.[1] The request must include the following information:

(1) A description of the licensing action with a citation to this Federal Register notice;

(2) The name and address of the potential party and a description of the potential party's particularized interest that could be harmed by the action identified in C.(1); and

(3) The identity of the individual or entity requesting access to SUNSI and the requester's basis for the need for the information in order to meaningfully participate in this adjudicatory proceeding. In particular, the request must explain why publicly-available versions of the information requested would not be sufficient to provide the basis and specificity for a proffered contention.

D. Based on an evaluation of the information submitted under paragraph C.(3) the NRC staff will determine within 10 days of receipt of the request whether:

(1) There is a reasonable basis to believe the petitioner is likely to establish standing to participate in this NRC proceeding; and

(2) The requestor has established a legitimate need for access to SUNSI.

E. If the NRC staff determines that the requestor satisfies both D.(1) and D.(2) above, the NRC staff will notify the requestor in writing that access to SUNSI has been granted. The written notification will contain instructions on how the requestor may obtain copies of the requested documents, and any other conditions that may apply to access to those documents. These conditions may include, but are not limited to, the signing of a Non-Disclosure Agreement or Affidavit, or Protective Order [2] setting forth terms and conditions to prevent the unauthorized or inadvertent disclosure of SUNSI by each individual who will be granted access to SUNSI.

F. Filing of Contentions. Any contentions in these proceedings that are based upon the information received as a result of the request made for SUNSI must be filed by the requestor no later than 25 days after the requestor is granted access to that information. However, if more than 25 days remain between the date the petitioner is granted access to the information and the deadline for filing all other contentions (as established in the notice of hearing or opportunity for hearing), the petitioner may file its SUNSI contentions by that later deadline. This provision does not extend the time for filing a request for a hearing and petition to intervene, which must comply with the requirements of 10 CFR 2.309.

G. Review of Denials of Access.

(1) If the request for access to SUNSI is denied by the NRC staff after a determination on standing and need for access, the NRC staff shall immediately notify the requestor in writing, briefly stating the reason or reasons for the denial.

(2) The requester may challenge the NRC staff's adverse determination by filing a challenge within 5 days of receipt of that determination with: (a) The presiding officer designated in this proceeding; (b) if no presiding officer has been appointed, the Chief Administrative Judge, or if he or she is unavailable, another administrative judge, or an administrative law judge with jurisdiction pursuant to 10 CFR 2.318(a); or (c) officer if that officer has been designated to rule on information access issues.

H. Review of Grants of Access. A party other than the requester may challenge an NRC staff determination granting access to SUNSI whose release would harm that party's interest independent of the proceeding. Such a challenge must be filed with the Chief Administrative Judge within 5 days of the notification by the NRC staff of its grant of access.

If challenges to the NRC staff determinations are filed, these procedures give way to the normal process for litigating disputes concerning access to information. The availability of interlocutory review by the Commission of orders ruling on such NRC staff determinations (whether granting or denying access) is governed by 10 CFR 2.311.[3]

I. The Commission expects that the NRC staff and presiding officers (and Start Printed Page 53464any other reviewing officers) will consider and resolve requests for access to SUNSI, and motions for protective orders, in a timely fashion in order to minimize any unnecessary delays in identifying those petitioners who have standing and who have propounded contentions meeting the specificity and basis requirements in 10 CFR Part 2. Attachment 1 to this Order summarizes the general target schedule for processing and resolving requests under these procedures.

It is so ordered.

Start Signature

Dated at Rockville, Maryland, this 14th day of August 2014.

For the Nuclear Regulatory Commission.

Annette L. Vietti-Cook,

Secretary of the Commission.

End Signature

Attachment 1—General Target Schedule for Processing and Resolving Requests for Access to Sensitive Unclassified Non-Safeguards Information in This Proceeding

DayEvent/activity
0Publication of Federal Register notice of hearing and opportunity to petition for leave to intervene, including order with instructions for access requests.
10Deadline for submitting requests for access to Sensitive Unclassified Non-Safeguards Information (SUNSI) with information: Supporting the standing of a potential party identified by name and address; describing the need for the information in order for the potential party to participate meaningfully in an adjudicatory proceeding.
60Deadline for submitting petition for intervention containing: (i) Demonstration of standing; and (ii) all contentions whose formulation does not require access to SUNSI (+25 Answers to petition for intervention; +7 petitioner/requestor reply).
20U.S. Nuclear Regulatory Commission (NRC) staff informs the requester of the staff's determination whether the request for access provides a reasonable basis to believe standing can be established and shows need for SUNSI. (NRC staff also informs any party to the proceeding whose interest independent of the proceeding would be harmed by the release of the information.) If NRC staff makes the finding of need for SUNSI and likelihood of standing, NRC staff begins document processing (preparation of redactions or review of redacted documents).
25If NRC staff finds no “need” or no likelihood of standing, the deadline for petitioner/requester to file a motion seeking a ruling to reverse the NRC staff's denial of access; NRC staff files copy of access determination with the presiding officer (or Chief Administrative Judge or other designated officer, as appropriate). If NRC staff finds “need” for SUNSI, the deadline for any party to the proceeding whose interest independent of the proceeding would be harmed by the release of the information to file a motion seeking a ruling to reverse the NRC staff's grant of access.
30Deadline for NRC staff reply to motions to reverse NRC staff determination(s).
40(Receipt +30) If NRC staff finds standing and need for SUNSI, deadline for NRC staff to complete information processing and file motion for Protective Order and draft Non-Disclosure Affidavit. Deadline for applicant/licensee to file Non-Disclosure Agreement for SUNSI.
AIf access granted: Issuance of presiding officer or other designated officer decision on motion for protective order for access to sensitive information (including schedule for providing access and submission of contentions) or decision reversing a final adverse determination by the NRC staff.
A + 3Deadline for filing executed Non-Disclosure Affidavits. Access provided to SUNSI consistent with decision issuing the protective order.
A + 28Deadline for submission of contentions whose development depends upon access to SUNSI. However, if more than 25 days remain between the petitioner's receipt of (or access to) the information and the deadline for filing all other contentions (as established in the notice of hearing or opportunity for hearing), the petitioner may file its SUNSI contentions by that later deadline.
A + 53(Contention receipt +25) Answers to contentions whose development depends upon access to SUNSI.
A + 60(Answer receipt +7) Petitioner/Intervenor reply to answers.
>A + 60Decision on contention admission.
End Supplemental Information

Footnotes

1.  While a request for hearing or petition to intervene in this proceeding must comply with the filing requirements of the NRC's “E-Filing Rule,” the initial request to access SUNSI under these procedures should be submitted as described in this paragraph.

Back to Citation

2.  Any motion for Protective Order or draft Non-Disclosure Affidavit or Agreement for SUNSI must be filed with the presiding officer or the Chief Administrative Judge if the presiding officer has not yet been designated, within 30 days of the deadline for the receipt of the written access request.

Back to Citation

3.  Requesters should note that the filing requirements of the NRC's E-Filing Rule (72 FR 49139; August 28, 2007) apply to appeals of NRC staff determinations (because they must be served on a presiding officer or the Commission, as applicable), but not to the initial SUNSI request submitted to the NRC staff under these procedures.

Back to Citation

[FR Doc. 2014-19880 Filed 9-8-14; 8:45 am]

BILLING CODE 7590-01-P