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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.

Enhanced Content

Relevant information about this document from Regulations.gov provides additional context. This information is not part of the official Federal Register document.

Published Document

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

ACTION:

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

DATES:

Comments must be filed by July 5, 2012. A request for a hearing or leave to intervene must be filed by August 6, 2012. Any potential party as defined in Title 10 of the Code of Federal Regulations (10 CFR) 2.4, who believes access to Sensitive Unclassified Non-Safeguards Information (SUNSI) is necessary to respond to this notice must request document access by June 15, 2012.

ADDRESSES:

You may access information and comment submissions related to this document, which the NRC possesses and are publicly available, by searching on http://www.regulations.gov under Docket ID NRC-2012-0125. You may submit comments by the following methods:

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0125. Address questions about NRC dockets to Carol Gallagher; telephone: 301-492-3668; email: Carol.Gallagher@nrc.gov.
  • Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
  • Fax comments to: RADB at 301-492-3446.

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

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2012-0125 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-2012-0125.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced in this notice (if that document is available in ADAMS) is provided the first time that a document is referenced.
  • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2012-0125 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 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.

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

This notice includes notices of amendments containing SUNSI.

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; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

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

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC'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 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, 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.

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

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

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

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

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

A person filing electronically using the agency's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC 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 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. 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. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

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

Date of amendment request: February 28, 2012. Publicly available versions of the amendment request and its attachment are available in ADAMS under Accession Nos. ML12061A288 and ML12061A289.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would revise the Palisades Nuclear Plant Technical Specifications (TS) to support the replacement of the Region I spent fuel pool (SFP) storage racks with new neutron absorber Metamic-equipped racks. Degradation of the present neutron absorber, Carborundum®, has resulted in reduced SFP storage capacity. The replacement of the SFP storage racks will allow recovery of the currently unusable storage locations in the SFP.

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 probability of any accident previously evaluated is not significantly increased by the proposed changes to the Region I spent fuel pool (SFP) storage racks. The probabilities of an accidental fuel assembly drop or misloading is primarily influenced by the methods used to lift and move these loads. The method of handling fuel is not changed since the same equipment and procedures will be used. Work in the SFP area will be controlled and performed in accordance with written procedures. Any movement of fuel assemblies required to be performed to support the modification will be performed in the same manner as during normal operations. Replacing the Region I SFP storage racks does not have a significant impact on the frequency of occurrence for any accident previously evaluated. Additionally, the probabilities of a seismic event, boron dilution, or loss of SFP cooling flow are not influenced by the proposed changes. Therefore, the proposed change will not involve a significant increase in the probability of occurrence of any event previously analyzed.

TS 3.7.15, SFP Boron Concentration, requires a minimum boron concentration of 1720 ppm, which bounds the analysis for the proposed amendment. Soluble boron is maintained in the SFP water as required by the TS and controlled by procedures. The criticality safety analyses for Region I and for Region II of the SFP credit the same soluble boron concentration of 850 ppm to maintain a Keff ≤ 0.95 under normal conditions and 1350 ppm to maintain a Keff ≤ 0.95 under accident scenarios as does the analysis for the proposed change for Region I. In crediting soluble boron, in Region 1, the SFP criticality analysis would have no effect on normal pool operation and maintenance. Thus, there is no change to the probability or the consequences of the boron dilution event in the SFP.

The consequences of the dropped spent fuel assembly in the SFP have been re-evaluated for the proposed change by analyzing a potential impact on the replacement racks. The results show that the postulated accident of a fuel assembly striking the top of the replacement racks would not distort the racks sufficiently to impair their functionality. The minimum subcriticality margin (i.e., neutron multiplication factor (Keff) less than or equal to 0.95) will be maintained. The structural damage to the fuel building, pool liner, and fuel assembly resulting from a dropped fuel assembly striking the pool floor or another assembly located in the racks is primarily dependent on the mass of the falling object and drop height. Since these two parameters are not changed by the proposed modification, the postulated structural damage to these items remains unchanged. The radiological dose at the exclusion area boundary has been evaluated and found to remain well below levels established by regulatory guidance.

The consequences of a loss of SFP cooling were evaluated and found to not involve a significant increase as a result of the proposed changes. The concern with this accident is a reduction of SFP water inventory from bulk pool boiling resulting in uncovering fuel assemblies. This situation could lead to fuel failure and subsequent significant increase in offsite dose. Loss of SFP cooling is mitigated by ensuring that a sufficient time lapse exists between the loss of forced cooling and the uncovering of fuel. This period of time is compared against a reasonable period to re-establish cooling or supply an alternative water source. Evaluation of this accident includes determination of the time to boil. This time period is much less than the onset of any significant increase in offsite dose, since once boiling begins it would have to continue unchecked until the pool water surface was lowered to the point of exposing active fuel. The time to boil represents the onset of loss of pool water inventory and is used as a gauge for establishing the comparison of consequences before and after a rack replacement project. The heatup rate in the SFP is a nearly linear function of the fuel decay heat load. The thermal-hydraulic analysis determined that the minimum time to boil is at least 1.8 hours subsequent to complete loss of forced cooling. In the unlikely event that all pool cooling is lost, sufficient time will still be available subsequent to the proposed changes for the operators to provide alternate means of cooling before the water shielding above the top of the racks falls below an acceptable level.

The consequences of a design basis seismic event are not increased. The consequences of this event are evaluated on the basis of subsequent fuel damage or compromise of the fuel storage or building configurations leading to radiological or criticality concerns. The replacement racks have been analyzed and were found to be safe during seismic motion. Fuel has been determined to remain intact and the storage racks maintain the fuel and fixed neutron absorber configurations subsequent to a seismic event. The replacement racks do not impact the pool walls. The structural capability of the pool and liner will not be exceeded under the appropriate combinations of dead weight, thermal, and seismic loads. The fuel building structure will remain intact during a seismic event and will continue to adequately support and protect the fuel racks, storage array, and pool moderator and coolant.

The consequence of a fuel misloading accident has been analyzed for the worst possible storage configuration subsequent to the proposed modification and it has been shown that the consequences remain acceptable with respect to the same criteria used previously.

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

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

Response: No.

The existing TS allow storage of fuel assemblies with a maximum planar average U-235 enrichment of 4.54 weight percent. The proposed change, for the replacement Region I fuel storage racks, would increase the maximum planar average U-235 enrichment to 4.95 weight percent. Fuel would be allowed in all the storage cell locations in the Metamic TM equipped Region I storage racks. Therefore, the possibility of misplacing a fuel assembly in the replacement fuel storage racks, with greater enrichment than allowed in certain storage locations in Region I, for the Carborundum equipped fuel storage racks would be eliminated, for the replacement Metamic TM equipped fuel storage racks. Changing the enrichment and allowing fuel storage in all the storage locations in the Metamic TM equipped Region I storage racks does not create a new or different kind of accident from any previously evaluated.

No new or different activities are introduced in the replacement of the fuel storage racks other than the physical removal of the existing racks and the installation of the new Metamic TM equipped fuel storage racks. An accident of a rack dropping onto stored spent fuel or the pool floor liner is not a postulated event due to the defense-in-depth approach to be taken. A rack lifting rig will be introduced to remove the existing Region I racks and to install the replacement racks. The temporary lift items are designed to meet the requirements of NUREG-0612, Control of Heavy Loads, and ANSI N14.6, Radioactive Materials—Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More. The lift rig and rack would be lifted using the fuel building crane main hook. This crane and main hook satisfy the single failure proof criteria of NUREG-0554, Single Failure Proof Cranes for Nuclear Power Plants. A rack drop event is a “heavy load drop” over the SFP. A lifted rack will not be allowed to travel over any stored fuel assemblies, thus a rack drop onto fuel is precluded. A rack drop to the pool liner is not a postulated event. All movements of heavy loads over the pool will comply with the applicable administrative controls and guidelines. Therefore, the activities for removal and installation of the fuel storage racks will not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed change does not alter the operating requirements of the plant or of the equipment credited in the mitigation of the design basis accidents. The changes would not affect the SFP cooling system or the SFP makeup systems. The proposed change does not affect any of the important parameters required to ensure safe fuel storage.

In summary, 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 function of the SFP is to store the fuel assemblies in a subcritical and cool-able configuration through all environmental and abnormal loadings, such as an earthquake or fuel assembly drop. The replacement Region I fuel storage rack design must meet all applicable requirements for safe storage and be functionally compatible with the SFP.

Detailed analysis has shown, with a 95 percent probability at a 95 percent confidence level, that the Keff of the Region I fuel storage racks in the SFP, including uncertainties, is less than 1.0 with unborated water and is less than or equal to 0.95 with 850 ppm of soluble boron in the SFP. In addition, the effects of abnormal and accident conditions have been evaluated to demonstrate that under credible conditions the Keff will not exceed 0.95 with 1350 ppm soluble boron credited. The current TS requirement for minimum SFP boron concentration is 1720 ppm, which provides assurance that the SFP would remain subcritical under normal, abnormal, or accident conditions. The margin of safety for subcriticality is maintained by having Keff equal to or less than 0.95 under all normal storage, fuel handling, and accident conditions, including uncertainties.

The current analysis basis for the Region I and Region II fuel storage racks is a maximum Keff of less than 1.0 when flooded with unborated water, and less than or equal to 0.95 when flooded with water having a boron concentration of 850 ppm. In addition, the Keff in accident or abnormal operating conditions is less than 0.95 with 1350 ppm of soluble boron. These values are not affected by the proposed change. Therefore, the margin of safety is not reduced.

The mechanical, material, and structural designs of the replacement racks have been reviewed in accordance with the applicable NRC guidance. The rack materials used are compatible with the spent fuel assemblies and the SFP environment. The design of the replacement racks preserves the margin of safety during abnormal loads such as a dropped fuel assembly. It has been shown that such loads will not invalidate the mechanical design and material selection to safely store fuel in a cool-able and subcritical configuration.

The thermal-hydraulic and cooling evaluation of the pool demonstrated that the pool can be maintained below the specified thermal limits under the conditions of the maximum heat load and during all credible accident sequences and seismic events. The pool temperature will not exceed 150°F during the worst single failure of a cooling pump. The maximum local water and fuel cladding temperatures in the hot channel will remain below the boiling point. The fuel will not undergo any significant heat up after an accidental drop of a fuel assembly on top of the rack blocking the flow path. A loss of cooling to the pool will allow sufficient time (nearly 2 hours) for the operators to intervene and line up alternate cooling paths and the means of inventory make-up before boiling begins. The thermal limits specified for the evaluations performed to support the proposed change are the same as those that were used in the previous evaluations.

Therefore, the proposed change for the replacement of the Region I SFP storage racks 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 Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Acting Branch Chief: Istvan Frankl.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of application for amendments: November 3, 2011. A publicly available version is available in ADAMS under Accession No. ML113081441.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would modify the Technical Specifications to include the use of neutron absorbing spent fuel pool rack inserts for the purpose of criticality control in the spent fuel pools.

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 revises Technical Specification (TS) 4.3.1 to permit installation of NETCO-SNAP-IN® rack inserts in spent fuel storage rack cells. The change is necessary to ensure that, with continued Boraflex degradation over time, the effective neutron multiplication factor, Keff, is less than or equal to 0.95, if the spent fuel pool is fully flooded with unborated water as required by 10 CFR 50.68. Because the proposed change pertains only to the spent fuel pool, only those accidents that are related to movement and storage of fuel assemblies in the spent fuel pool could potentially be affected by the proposed change.

The installation of NETCO-SNAP-IN® rack inserts does not result in a significant increase in the probability of an accident previously analyzed because there are no changes in the manner in which spent fuel is handled, moved, or stored in the rack cells. The probability that a fuel assembly would be dropped is unchanged by the installation of the NETCO-SNAP-IN® rack inserts. These events involve failures of administrative controls, human performance, and equipment failures that are unaffected by the presence or absence of Boraflex and the rack inserts.

The installation of NETCO-SNAP-IN® rack inserts does not result in a significant increase in the consequences of an accident previously analyzed because there is no change to the fuel assemblies that provide the source term used in calculating the radiological consequences of a fuel handling accident. In addition, consistent with the current design, only one fuel assembly will be moved at a time. Thus, the consequences of dropping a fuel assembly onto any other fuel assembly or other structure remain bounded by the previously analyzed fuel handling accident. The proposed change does not affect the effectiveness of the other engineered design features to limit the offsite dose consequences of the limiting fuel handling accident.

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

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

Response: No.

Onsite storage of spent fuel assemblies in the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 spent fuel pools is a normal activity for which PBAPS has been designed and licensed. As part of assuring that this normal activity can be performed without endangering public health and safety, the ability to safely accommodate different possible accidents in the spent fuel pool, such as dropping a fuel assembly or misloading a fuel assembly, have been analyzed. The proposed spent fuel storage configuration does not change the methods of fuel movement or spent fuel storage. The proposed change allows for continued use of spent fuel pool storage rack cells with degraded Boraflex within those spent fuel pool storage rack cells.

The rack inserts are passive devices. These devices, when inside a spent fuel storage rack cell, perform the same function as the previously licensed Boraflex neutron absorber panels in that cell. These devices do not add any limiting structural loads or affect the removal of decay heat from the assemblies. No change in total heat load in the spent fuel pool is being made. The devices will maintain their design function over the life of the spent fuel pool. The existing fuel handling accident, which assumes the drop of a fuel assembly, bounds the drop of a rack insert and/or rack insert installation tool. This proposed change does not create the possibility of misloading an assembly into a spent fuel storage rack cell.

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.

PBAPS TS 4.3.1.1 requires the spent fuel storage racks to maintain the effective neutron multiplication factor, Keff, less than or equal to 0.95 when fully flooded with unborated water, which includes an allowance for uncertainties. Therefore, for criticality, the required safety margin is 5% including a conservative margin to account for engineering and manufacturing uncertainties.

The proposed change provides a method to ensure that Keff continues to be less than or equal to 0.95, thus preserving the required safety margin of 5%. The criticality analyses demonstrate that the required margin to criticality of 5%, including a conservative margin to account for engineering and manufacturing uncertainties, is maintained assuming an infinite array of fuel with all fuel at the peak reactivity. In addition, the radiological consequences of a dropped fuel assembly are unchanged because the event involving a dropped fuel assembly onto a spent fuel storage rack cell containing a fuel assembly with a rack insert is bounded by the radiological consequences of a dropped fuel assembly without a rack insert. The proposed change also maintains the capacity of the Unit 2 and Unit 3 spent fuel pools to be no more than 3,819 fuel assemblies each.

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. J. Bradley Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Meena K. Khanna.

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

Date of amendment request: April 10, 2012. A publically available version is available in ADAMS under Accession No. ML12121A527.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed changes would revise the Seabrook Station Technical Specifications (TSs). The proposed change would revise TS 6.7.6.k, Steam Generator (SG) Program, to exclude a portion of the tubes below the top of the SG tube sheet from periodic tube inspections and plugging. The proposed change also establishes permanent reporting requirements in TS 6.8.1.7, Steam Generator Tube Inspection Report, that were previously implemented on a temporary basis.

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

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

Response: No.

The previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed change that alters the steam generator (SG) inspection and reporting criteria does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident.

Of the applicable accidents previously evaluated, the limiting transients with consideration to the proposed change to the SG tube inspection and repair criteria are the steam generator tube rupture (SGTR) event, the steam line break (SLB), and the feed line break (FLB) postulated accidents.

Addressing the SGTR event, the required structural integrity margins of the SG tubes and the tube-to-tubesheet joint over the H* distance will be maintained. Tube rupture in tubes with cracks within the tubesheet is precluded by the constraint provided by the presence of the tubesheet and the tube-to-tubesheet joint. Tube burst cannot occur within the thickness of the tubesheet. The tube-to-tubesheet joint constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet, and from the differential pressure between the primary and secondary side, and tubesheet rotation. The structural margins against burst, as discussed in Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator Tubes,” and Technical Specification (TS) 6.7.6.k, are maintained for both normal and postulated accident conditions.

The proposed change has no impact on the structural or leakage integrity of the portion of the tube outside of the tubesheet. The proposed change maintains structural and leakage integrity of the SG tubes consistent with the performance criteria of TS 6.7.6.k. Therefore, the proposed change results in no significant increase in the probability of the occurrence of a SGTR accident.

At normal operating pressures, leakage from tube degradation below the proposed limited inspection depth is limited by the tube-to-tubesheet crevice. Consequently, negligible normal operating leakage is expected from degradation below the inspected depth within the tubesheet region. The consequences of an SGTR event are not affected by the primary-to-secondary leakage flow during the event as primary-to-secondary leakage flow through a postulated tube that has been pulled out of the tubesheet is essentially equivalent to a severed tube. Therefore, the proposed change does not result in a significant increase in the consequences of a SGTR.

The consequences of a SLB or FLB are also not significantly affected by the proposed changes. The leakage analysis shows that the primary-to-secondary leakage during a SLB/FLB event would be less than or equal to that assumed in the Updated Safety Analysis Report.

Primary-to-secondary leakage from tube degradation in the tubesheet area during the limiting accident (i.e., a SLB/FLB) is limited by flow restrictions. These restrictions result from the crack and tube-to-tubesheet contact pressures that provide a restricted leakage path above the indications and also limit the degree of potential crack face opening as compared to free span indications.

The leakage factor of 2.49 for Seabrook Station, for a postulated SLB/FLB, has been calculated as shown in References 8, 9 and 10. For the Condition Monitoring assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.49 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the Operational Assessment, the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.49 and compared to the observed operational leakage.

The probability of a SLB/FLB is unaffected by the potential failure of a SG tube as the failure of the tube is not an initiator for a SLB/FLB event. SLB/FLB leakage is limited by flow restrictions resulting from the leakage path above potential cracks through the tube-to-tubesheet crevice. The leak rate during all postulated accident conditions that model primary-to-secondary leakage (including locked rotor and control rod ejection) has been shown to remain within the accident analysis assumptions for all axial and or circumferentially orientated cracks occurring 15.21 inches below the top of the tubesheet. The assumed accident induced leak rate for Seabrook is 500 gallons per day (gpd) during a postulated steam line break in the faulted loop. Using the limiting leak rate factor of 2.49, this corresponds to an acceptable level of operational leakage of 200 gpd. Therefore, the TS leak rate limit of 150 gpd provides significant added margin against the 500 gpd accident analysis leak rate assumption.

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

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

Response: No.

The proposed change that alters the SG inspection and reporting criteria does not introduce any new equipment, create new failure modes for existing equipment, or create any new limiting single failures. Plant operation will not be altered, and all safety functions will continue to perform as previously assumed in accident analyses.

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

3. The proposed changes do not involve a significant reduction in the margin of safety.

Response: No.

The proposed change that alters the SG inspection and reporting criteria maintains the required structural margins of the SG tubes for both normal and accident conditions. Nuclear Energy Institute 97-06, Rev. 3 “Steam Generator Program Guidelines,” and NRC Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR Steam Generator Tubes,” are used as the bases in the development of the limited hot leg tubesheet inspection depth methodology for determining that SG tube integrity considerations are maintained within acceptable limits. RG 1.121 describes a method acceptable to the NRC for meeting General Design Criteria (GDC) 14, “Reactor Coolant Pressure Boundary,” GDC 15, “Reactor Coolant System Design,” GDC 31, “Fracture Prevention of Reactor Coolant Pressure Boundary,” and GDC 32, “Inspection of Reactor Coolant Pressure Boundary,” by reducing the probability and consequences of a SGTR. RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation, the probability and consequences of a SGTR are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.

For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially oriented cracking, Westinghouse WCAP-17071-P defines a length of degradation-free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary-to-secondary leakage during all plant conditions. The methodology for determining leakage as described in WCAP-17071-P provides significant margin between the accident-induced leakage assumption and the technical specification leakage limit during normal operating conditions when the proposed limited tubesheet inspection depth criteria is implemented.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC.

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

NRC Branch Chief: Meena Khanna.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, California

Date of amendment request: October 26, 2011. A publicly available version is available in ADAMS under Accession No. ML113070457.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendment would revise the facility operating licenses to allow the permanent replacement of the current Diablo Canyon Power Plant, Units 1 and 2 (DCPP) Eagle 21 digital process protection system (PPS) with a new digital PPS that is based on the Invensys Operations Management Tricon Programmable Logic Controller (PLC), Version 10, and the CS Innovations, LLC (CSI, a Westinghouse Electric Company), Advanced Logic System (ALS).

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

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

Response: No.

The proposed change would allow Pacific Gas and Electric Company to permanently replace the Diablo Canyon Power Plant Eagle 21 digital process protection system with a new digital process protection system that is based on the Invensys Operations Management Tricon Programmable Logic Controller, Version 10, and the CS Innovations Advanced Logic System. The process protection system replacement is designed to applicable codes and standards for safety-grade protection systems for nuclear power plants and incorporates additional redundancy and diversity features and therefore, does not result in an increase in the probability of inadvertent actuation or probability of failure to initiate a protective function. The process protection system replacement does not introduce any new credible failure mechanisms or malfunctions that cause an accident. The process protection system replacement design will continue to perform the reactor trip system and engineered safety features actuation system functions assumed in the Final Safety Analysis Report within the response time assumed in the Final Safety Analysis Report Chapter 6 and 15 accident analyses.

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 accident from any accident previously evaluated?

Response: No.

The proposed change is to permanently replace the current Diablo Canyon Power Plant Eagle 21 digital process protection system with a new digital process protection system. The process protection system performs the process protection functions for the reactor protection system that monitors selected plant parameters and initiates protective action as required. Accidents that may occur due to inadvertent actuation of the process protection system, such as an inadvertent safety injection actuation, are considered in the Final Safety Analysis Report accident analyses.

The protection system is designed with redundancy such that a single failure to generate an initiation signal in the process protection system will not cause failure to trip the reactor nor failure to actuate the engineered safeguard features when required. Neither will such a single failure cause spurious or inadvertent reactor trips or engineered safeguard features actuations because coincidence of two or more initiation signals is required for the solid state protection system to generate a trip or actuation command. If an inadvertent actuation occurs for any reason, existing control room alarms and indications will notify the operator to take corrective action.

The process protection system replacement design includes enhanced diversity features compared to the current process protection system to provide additional assurance that the protection system actions credited with automatic operation in the Final Safety Analysis Report accident analyses will be performed automatically when required should a common cause failure occur concurrently with a design basis event.

The process protection system replacement does not result in any new credible failure mechanisms or malfunctions. The current Eagle 21 process protection system utilizes digital technology and therefore the use of digital technology in the process protection system replacement does not introduce a new type of failure mechanism. Although extremely unlikely, the current Eagle 21 process protection system is susceptible to a credible common-cause software failure that could adversely affect automatic performance of the protection function. The process protection system replacement contains new, additional diversity features that prevent a common-cause software failure from completely disabling the process protection system.

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

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

Response: No.

The reactor protection system is fundamental to plant safety and performs reactor trip system and engineered safety features actuation system functions to limit the consequences of Condition II (faults of moderate frequency), Condition III (infrequent faults), and Condition IV (limiting faults) events. This is accomplished by sensing selected plant parameters and determining whether predetermined instrument settings are being exceeded. If predetermined instrument settings are exceeded, the reactor protection system sends actuation signals to trip the reactor and actuate those components that mitigate the severity of the accident.

The process protection system replacement design will continue to perform the reactor trip system and engineered safety features actuation functions assumed in the Final Safety Analysis Report within the response time assumed Final Safety Analysis Report Chapter 6 and 15 accident analyses. The use of the process protection system replacement does not result in a design basis or safety limit being exceeded or changed. The change to the process protection system has no impact on the reactor fuel, reactor vessel, or containment fission product barriers. The reliability and availability of the reactor protection system is improved with the process protection system replacement, and the reactor protection system will continue to effectively perform its function of sensing plant parameters to initiate protective actions to limit or mitigate events.

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: Jennifer Post, Esq., Pacific Gas and Electric Company, 77 Beale Street, Room 2496, Mail Code B30A, San Francisco, CA 94105.

NRC Branch Chief: Michael T. Markley.

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

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

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania

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

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, California

A. This Order contains instructions regarding how potential parties to this proceeding may request access to documents containing Sensitive Unclassified Non-Safeguards Information (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 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 requestor 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 requestor'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.

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 requestor 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) if another officer has been designated to rule on information access issues, with that officer.

H. Review of Grants of Access. A party other than the requestor 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 any 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.

Dated at Rockville, Maryland, this 29th day of May, 2012.

For the Nuclear Regulatory Commission.

Annette L. Vietti-Cook,

Secretary of the Commission.

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; (ii) all contentions whose formulation does not require access to SUNSI (+25 Answers to petition for intervention; +7 requestor/petitioner reply).
20Nuclear Regulatory Commission (NRC) staff informs the requestor 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 requestor/petitioner 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.

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.

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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.

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3.  Requestors 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.

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[FR Doc. 2012-13426 Filed 6-4-12; 8:45 am]

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