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

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Information about this document as published in the Federal Register.

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I. 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 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 sensitive unclassified non-safeguards information (SUNSI) or safeguards information (SGI).

Notice of Consideration of Issuance of Amendments to Facility Operating 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 Start Printed Page 4248no 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.

Written comments may be submitted by mail to the Chief, Rulemaking, Directives and Editing Branch, TWB-05-B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Documents may be examined, and/or copied for a fee, at the NRC(s Public Document Room (PDR), located at One White Flint North, Public File Area O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland.

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. 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 Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, or at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​part002/​part002-0309.html. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm.html. 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 petitioner/requestor 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 petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor 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, 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, which the NRC promulgated in August 28, 2007 (72 FR 49139). 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 Start Printed Page 4249a waiver in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by calling (301) 415-1677, to request (1) a digital 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/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms Viewer(tm) to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms Viewer(tm) is free and is available at http://www.nrc.gov/​site-help/​e-submittals/​install-viewer.html. Information about applying for a digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.html.

Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE 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 e-mail notice confirming receipt of the document. The EIE system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically may seek assistance through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html or by calling the NRC electronic filing Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday. The electronic filing Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at MSHD.Resource@nrc.gov.

Participants who believe that they have a good cause for not submitting documents electronically must file a motion, 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.

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/​ehd_​proceeding/​home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a 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. 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.

For further details with respect to this amendment action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, 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 Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr.resource@nrc.gov.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: November 20, 2008.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment revises Technical Specification (TS) 5.6.5, “Core Operating Limits Report (COLR),” to add a reference to an analytical method that will be used to determine core operating limits. The new reference, NEDC-33383P, “GEXL97 Correlation Applicable to ATRIUM-10 Fuel,” will allow the licensee to use a Global Nuclear Fuel method to determine fuel assembly critical power of AREVA ATRIUM-10 fuel.

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.

Core operating limits are established each operating cycle in accordance with TS 3.2, “Power Distribution” and TS 5.6.5, “Core Operating Limits Report (COLR).” These core operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO). The methods used to determine the operating limits are those previously found acceptable by the NRC and listed in TS section 5.6.5.b.

A change to TS 5.6.5.b is requested to include an additional reference to the list of analytical methods. RBS [River Bend Station] currently operates with a full core of AREVA ATRIUM-10 fuel but is scheduled to load GE14 fuel during the next refueling outage. Start Printed Page 4250RBS plans to use the analysis methods of the new fuel vendor, GNF [Global Nuclear Fuel], for the analysis of the mixed core. The GEXL97 correlation accurately models predicted core behavior and appropriately determines the overall critical power uncertainty of this method. In addition, the GEXL97 application range covers the range of expected operation of the ATRIUM-10 fuel during normal steady state and transient conditions in the RBS reload cores.

The requested TS changes concern the use of analytical methods and do not involve any plant modifications or operational changes that could affect any postulated accident precursors or accident mitigation systems and do not introduce any new accident initiation mechanisms. The proposed changes have no effect on the type or amount of radiation released and [have] no effect on predicted offsite doses in the event of an accident. Thus, the proposed change does not affect the probability of an accident previously evaluated nor does it increase the radiological consequences of any accident previously evaluated.

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

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

Response: No.

The proposed TS changes will not change the design function, reliability, performance, or operation of any plant systems, components, or structures. It does not create the possibility of a new failure mechanism, malfunction, or accident initiators not considered in the design and licensing bases. Plant operation will continue to be within the core operating limits that are established using NRC approved methods that are applicable to the RBS design and the RBS fuel.

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 proposed change adds GEXL97 to the list of analytical methods in TS 5.6.5.b that can be used to determine core operating limits. Use of the GEXL97 correlation analytical method provides an equivalent level of protection as that currently provided. The change does not alter any method of analysis as described in the NRC approved versions of GESTAR-II [NEDE-24011-P-A, “General Electric Standard Application for Reactor Fuel (GESTAR-II)”]. The proposed change does not modify the safety limits or setpoints at which protective actions are initiated, and do not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the 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: Terence A. Burke, Associate General Counsel—Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, Mississippi 39213.

NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: December 16, 2008.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). This amendment request would revise the Technical Specifications (TSs) Section 2.1.2, Safety Limit Minimum Critical Power Ratio (SLMCPR) for two-loop and single-loop operation.

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

1. The proposed changes to Technical Specification do not involve a significant increase in the probability of an accident previously evaluated.

The proposed Safety Limit MCPR (SLMCPR), and its use to determine the Operating Cycle 18 thermal limits, have been derived using NRC approved methods specified in the Reference section of the Technical Specification Bases Section for 2.0 SAFETY LIMITS. These methods do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient.

The basis of the SLMCPR is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. The new SLMCPR preserves the margin to transition boiling, and the probability of fuel damage is not increased.

Therefore, the proposed changes to Technical Specifications do not involve an increase in the probability or consequences of an accident previously evaluated.

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

The proposed changes result only from revised methods of analysis for the Cycle 18 core reload. These methods have been reviewed and approved by the NRC, do not involve any new or unapproved method for operating the facility, and do not involve any facility modifications. No new initiating events or transients result from these changes.

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

3. The proposed changes to Technical Specifications do not involve a significant reduction in a margin of safety.

The margin of safety as defined in the TS bases will remain the same. The new SLMCPR was derived using NRC approved methods which are in accordance with the current fuel design and licensing criteria. The SLMCPR remains high enough to ensure that greater than 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity.

Therefore, the proposed changes to technical specifications do not involve a significant reduction in the margin of safety.

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

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

NRC Branch Chief: Mark G. Kowal.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: July 25, 2008.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendments would revise Technical Specification 3.3.1.1, “Reactor Protection System (RPS) Instrumentation,” Surveillance Requirement (SR) 3.3.1.1.8 and TS 3.3.1.3, “Oscillation Power Range Monitor (OPRM) Instrumentation,” SR 3.3.1.3.2 to increase the frequency interval between local power range monitor calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH for the LaSalle County Station, Units 1 and 2 (LSCS).

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?Start Printed Page 4251

Response: No.

The proposed change is a result of increasing the surveillance interval of the LPRM [Local Power Range Monitor] calibration frequency from 1000 EFPH to 2000 EFPH. Increasing the frequency interval between required LRPM calibrations is acceptable due to improvements in the fuel analytical bases and therefore, the revised surveillance interval continues to ensure that the LPRM detector signal is adequately calibrated. Extending the LPRM calibration surveillance interval will increase the LPRM signal uncertainty value used in the LSCS SLMCPR [Safety Limit Minimum Critical Power Ratio] analysis, however, this increase in the LRPM signal uncertainty value is acceptable since the increase is bounded by the values used by the AREVA analysis.

This change will not alter the operation of process variables, structures, systems, or components as described in the LSCS Updated Final Safety Analysis Report (UFSAR). The proposed change does not alter the initiation conditions or operational parameters for the system and there is no new equipment introduced by the extension of the LPRM calibration frequency interval. The performance of the Average Power Range Monitor (APRM), Rod Block Monitor (RBM) and Oscillation Power Range Monitor (OPRM) systems are not significantly affected by the proposed surveillance interval increase. The proposed LPRM calibration interval extension will have no significant effect on the Reactor Protection System (RPS) instrumentation accuracy during power maneuvers or transients and will, therefore, not significantly affect the performance of the RPS. As such, the probability of occurrences for a previously evaluated accident is not increased.

The radiological consequences of an accident can be affected by the thermal limits existing at the time of the postulated accident, however, increasing the surveillance interval frequency will not increase the calculated thermal limits since all uncertainties associated with the increased interval are currently implemented and are currently used to calculate the existing Safety Limits. Plant specific evaluation of LPRM sensitivity to exposure has determined that the extended calibration frequency increases the LPRM signal uncertainty value used in the LSCS SLMCPR analysis, however, the increase is bounded by the values currently used in the safety analysis. Therefore, the thermal limit calculation is not significantly affected by LPRM calibration frequency, and thus the radiological consequences of any accident previously evaluated are not increased.

Based on the above information, 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 performance of the APRM, RBM, and OPRM systems are not significantly affected by the proposed LPRM surveillance interval increase. The proposed change does not affect the control parameters governing unit operation or the response of plant equipment to transient conditions. For the proposed LPRM extended calibration interval frequency all uncertainties remain less than the uncertainties assumed in the existing thermal limit calculations. The proposed change does not change or introduce any new equipment, modes of system operation or failure mechanisms.

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

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

Response: No.

The proposed change has no impact on equipment design or fundamental operation, and there are no changes being made to safety limits or safety system allowable values that would adversely affect plant safety as a result of the proposed LPRM surveillance interval increase. The performance of the APRM, RBM, and OPRM systems are not significantly affected by the proposed change. The margin of safety can be affected by the thermal limits existing at the time of the postulated accident; however, uncertainties associated with LPRM chamber exposure have no significant effect on the calculated thermal limits. Plant specific evaluation of LPRM sensitivity to exposure has determined that the extended calibration frequency increases the LPRM signal uncertainty value used in the LSCS SLMCPR analysis, however, the increase is bounded by the values currently used in the safety analysis. The thermal limit calculation is not significantly affected since the LPRM sensitivity with exposure is well defined. LPRM accuracy remains within the total nodal power uncertainty assumed in the thermal analysis basis, therefore maintaining thermal limits and the safety margin. The proposed change does not affect safety analysis assumptions or initial conditions and the margin of safety in the original safety analysis are therefore maintained.

Based on this information, 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 requested amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell Gibbs.

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

Date of amendment request: October 13, 2008.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would revise the licensing basis by approving adoption of the Alternative Source Term (AST), in accordance with 10 CFR 50.67, for use in calculating the loss-of-coolant accident (LOCA) dose consequences. The proposed amendment would revise the Technical Specifications (TSs) to (1) change the TS definition for DOSE EQUIVALENT I-131 to adopt Federal Guidance Report (FGR) 11 dose conversion factors, (2) require operability of the Standby Liquid Control (SLC) system in Mode 3, to reflect its credit in the LOCA analysis, (3) establish a Main Steam (MS) Pathway leakage limit that effectively increases the previous MS isolation valve leakage limit, and (4) change TS Section 5.5.12 to reflect a requested permanent exemption from the requirements of 10 CFR Part 50, Appendix J, Option B, Paragraph III.A, to allow exclusion of MS Pathway leakage from the overall integrated leakage rate measured during the performance of a Type A test, and from the requirements of Appendix J, Option B, Paragraph III.B, to allow exclusion of the MS Pathway leakage from the combined leakage rate of the penetrations and valves subject to Type B and C tests.

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.

For the postulated design basis accident (DBA) LOCA, the AST is an input to the calculations that evaluate the radiological consequences of a LOCA. The AST and the requested Appendix J exemption do not affect the design of the plant or the manner in which the plant is normally operated. Adoption of the AST and the requested Appendix J exemption do not affect the initiators of a DBA. Neither the AST nor the requested Appendix J exemption [sic] affect the response to the DBA LOCA, or the pathway of the radiation released from the nuclear fuel. Rather, the AST better represents the physical characteristics of the radiation release.

Because the initiators of a DBA are not affected by adoption of the AST for LOCA dose assessment, the probability of an accident are not increased by the proposed amendment or requested Appendix J exemption.

The AST is an input to calculations used to evaluate the radiological consequences of Start Printed Page 4252the LOCA. Use of the AST does not affect the plant response to the accident, or the pathways to the environment for the radiation and activity released from the fuel. The LOCA radiological analyses have been performed using the AST. Adoption of the AST methodology revises the acceptance criteria for the accident to the limits specified in 10 CFR 50.67. The results of those analyses demonstrate that the dose consequences are within the acceptance criteria presented in 10 CFR 50.67 and in NRC RG [Regulatory Guide] 1.183.

Implementation of the AST for the LOCA involves the use of the SLC System to control the pH of the suppression pool during mitigation of a LOCA. As a result the proposed amendment revises the CNS [Cooper Nuclear Station] TS for the SLC System. These changes do not require any physical modification of the plant, nor result in any change in normal plant operation. This additional use of the SLC system does not compromise or adversely affect the function of the SLC system as a means of shutting down the reactor in addition to the control rods.

Therefore, it is concluded that adoption of AST and granting of the Appendix J exemption do not involve a significant increase in the consequences of an accident previously evaluated. Based on the above discussion, it is concluded that 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.

Implementation of the LOCA AST and the requested Appendix J exemption do not involve a physical alteration of the plant or a change in how the plant is normally operated. No new or different types of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed changes. The proposed changes, effectively increasing the allowable MSIV leakage, establishing a leakage limit for the MS Pathway, and crediting the SLC system for LOCA mitigation do not create initiators or precursors of a new or different kind of accident. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed amendment.

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

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

Response: No.

The proposed amendment involves the implementation of a new licensing basis for the design basis LOCA. Approval of this change from the original source term to an AST, derived in accordance with the guidance of RG 1.183, results in revised acceptance criteria for the LOCA analysis. For the LOCA, RG 1.183 sets the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room limit consistent with 10 CFR 50.67. The AST LOCA radiological analysis has been performed using conservative methodologies, as specified in RG 1.183. Safety margins have been evaluated and confirmed to have not been reduced. Analytical conservatism has been utilized to ensure that the analysis adequately bounds the limiting postulated event. The dose consequences of the DBA LOCA remain within the acceptance criteria presented in 10 CFR 50.67 and RG 1.183.

The proposed changes continue to ensure that the doses at the EAB and LPZ boundary, as well as the Control Room, are within the corresponding regulatory limits.

Since the proposed amendment continues to ensure the doses at the EAB, LPZ and Control Room are within corresponding regulatory limits, the proposed license amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Michael T. Markley.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: November 5, 2008.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The licensee proposed to increase the current maximum power level authorized by section 2.C(1) of the renewed facility operating license from 1,775 megawatts thermal (Mwt) to 2,004 Mwt, an approximately 13 percent increase from the current licensed thermal power. The current maximum power level of 1,775 Mwt was approved in 1998, an increase of 6.3 percent from the original licensed thermal power of 1670 Mwt. Thus, when approved, the licensee's proposed amendment would take the maximum power level to about 20 percent above the original license thermal power. The licensee's application addresses in details each of the following major technical areas: extended power uprate, containment analysis methods change, credit for containment overpressure for low head emergency core cooling system (ECCS) pumps, and reactor internal pressure differentials (RIPDs) for the steam dryer.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (10 CFR) Part 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration (NSHC). The licensee's NSHC analysis, addressing each technical area listed above, is reproduced below:

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

Extended Power Uprate

Response: No.

The probability (frequency of occurrence) of [d]esign [b]asis [a]ccidents occurring is not affected by the increased power level, because Monticello Nuclear Generating Plant (MNGP) continues to comply with the regulatory and design basis criteria established for plant equipment. A probabilistic risk assessment demonstrates that the calculated core damage frequencies do not significantly change due to [e]xtended [p]ower [u]prate (EPU). Scram setpoints (equipment settings that initiate automatic plant shutdowns) are established such that there is no significant increase in scram frequency due to EPU. No new challenges to safety-related equipment result from EPU.

The changes in consequences of postulated accidents, which would occur from 102 percent of the EPU rated thermal power (RTP) compared to those previously evaluated, are acceptable. The results of EPU accident evaluations do not exceed the NRC[-]approved acceptance limits. The spectrum of postulated accidents and transients has been investigated, and are shown to meet the plant's currently licensed regulatory criteria. In the area of fuel and core design, for example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and other applicable Specified Acceptable Fuel Design Limits (SAFDL) are still met. Continued compliance with the SLMCPR and other SAFDLs will be confirmed on a cycle[-]specific basis consistent with the criteria accepted by the NRC.

Challenges to the [r]eactor [c]oolant [p]ressure [b]oundary were evaluated at EPU conditions (pressure, temperature, flow, and radiation) and were found to meet their acceptance criteria for allowable stresses and overpressure margin.

Challenges to the containment have been evaluated, and the containment and its associated cooling systems continue to meet the current licensing basis. The increase in the calculated post[-] LOCA suppression pool temperature above the currently assumed peak temperature was evaluated and determined to be acceptable. Radiological release events (accidents) have been evaluated, and have been shown to meet the guidelines of 10 CFR 50.67.

Containment Analysis Methods Change

Response: No.

The use of passive heat sinks, variable RHR [residual heat removal] heat exchanger capability K-value, and mechanistic heat and Start Printed Page 4253mass transfer from the suppression pool surface to the wetwell airspace after 30 seconds for the long[-]term design[-] basis[-]accident loss[-]of[-]coolant accident (DBA-LOCA) containment analysis are not relevant to accident initiation, but rather, pertain to the method used to accurately evaluate postulated accidents. The use of these elements does not, in any way, alter existing fission product boundaries, and provides a conservative prediction of the containment response to DBA-LOCAs. Therefore, the containment analysis method change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Credit for Containment Overpressure for Low Head Emergency Core Cooling System (ECCS) Pumps

Response: No.

These changes update parameters used in the MNGP safety analyses and expand the range and scope of the analyses. This will result in a more realistic analysis of available containment overpressure under design[-]basis accident conditions. The updated analyses affect only the evaluation of previously reviewed accidents. No plant structure, system, or component (SSC) is physically affected by the updated and expanded analyses. No method of operation of any plant SSC is affected. Therefore, there is no significant increase in the probability or consequence of a previously evaluated accident.

Reactor Internal Pressure Differentials (RIPDs) for the Steam Dryer

Response: No.

The revised steam dryer RIPDs are used in evaluating loads in reactor vessel internals for various conditions (i.e., during normal, upset and faulted conditions). The values more accurately represent the actual plant configuration. No plant structure, system, or component (SSC) is physically affected by the updated and expanded analyses. No method of operation of any plant SSC is affected. Therefore, there is no significant increase in the probability or consequence of a previously evaluated accident.

The analyses supporting the above evaluations were performed at the EPU power level of 2,004 Mwt.

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?

Extended Power Uprate

Response: No.

Equipment that could be affected by EPU has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified. The full spectrum of accident considerations has been evaluated and no new or different kind of accident has been identified. EPU uses developed technology and applies it within capabilities of existing or modified plant safety[-]related equipment in accordance with the regulatory criteria (including NRC[-]approved codes, standards and methods). No new accidents or event precursors have been identified.

The MNGP TS require revision to implement EPU. The revisions have been assessed and it was determined that the proposed change will not introduce a different accident than that previously evaluated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Containment Analysis Methods Change

Response: No.

The use of passive heat sinks, variable RHR heat exchanger capability K-value, and mechanistic heat and transfer from the suppression pool surface to the wetwell airspace after 30 seconds for the long[-]term DBA-LOCA containment analysis are not relevant to accident initiation, but pertain to the method used to evaluate currently postulated accidents. The use of these analytical tools does not involve any physical changes to plant structures or systems, and does not create a new initiating event for the spectrum of events currently postulated. Further, they do not result in the need to postulate any new accident scenarios. Therefore, the containment analysis method change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Credit for Containment Overpressure for Low Head ECCS Pumps

Response: No.

The proposed change involves the updating and expansion in scope of the existing design bases analysis with respect to the available containment overpressure to cover additional events. No new failure mode or mechanisms have been created for any plant SSC important to safety nor has any new limiting single failure been identified as a result of the proposed analytical changes. Therefore, the change to containment overpressure credited for low pressure ECCS pumps does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Reactor Internal Pressure Differentials for the Steam Dryer

Response: No.

The revised steam dryer RIPDs are used in evaluating loads in reactor vessel internals for various conditions (i.e., during normal, upset and faulted conditions). The steam dryer RIPDs are not relevant to accident initiation, but only pertain to the method used to evaluate reactor vessel internals loads. The revised steam dryer RIPD values more accurately represent the actual plant configuration. Therefore, the change to steam dryer RIPDs does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The analyses supporting the above evaluations were performed at the EPU power level of 2,004 Mwt.

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

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

Extended Power Uprate

Response: No.

The EPU affects only design and operational margins. Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for EPU conditions. Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on affected structures, systems and components, including the reactor coolant pressure boundary, will remain within their design allowables for design[-]basis event categories. No NRC acceptance criterion is exceeded. Because the MNGP configuration and responses to transients and postulated accidents do not result in exceeding the presently approved NRC acceptance limits, the proposed changes do not involve a significant reduction in a margin of safety.

Containment Analysis Methods Change

Response: No.

The use of passive heat sinks, variable RHR heat exchanger capability K-value, and mechanistic heat and mass transfer from the suppression pool surface to the wetwell airspace after 30 seconds for the long[-]term DBA-LOCA containment analysis are realistic phenomena and provide a conservative prediction of the plant response to DBA-LOCAs. The increase in pressure and temperature are relatively small and are within design limits. Therefore, the containment analysis methods change does not involve a significant reduction in the margin of safety.

Increase in Credit for Containment Overpressure for Low Head ECCS Pumps

Response: No.

The proposed changes revise containment response analytical methods and scope for containment pressure to assist in ECCS pump net positive suction head (NPSH). The changes are still based on conservative but more realistic analysis of available containment overpressure determined using analysis methods that minimize containment pressure and maximize suppression pool temperature. These changes do not constitute a significant reduction in the margin of safety.

Reactor Internal Pressure Differentials for the Steam Dryer

Response: No.

The revised steam dryer RIPDs are used in evaluating loads in reactor vessel internals for various conditions (i.e., during normal, upset and faulted conditions). The revised steam dryer RIPD values more accurately represent the actual plant configuration. The changes are still conservative but more accurately represent the MNGP configuration. These changes do not constitute a significant reduction in the margin of safety.

The analyses supporting the above evaluations were performed at the EPU power level of 2,004 Mwt.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Start Printed Page 4254

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 proposed amendment involves 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: Lois M. James.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of amendment request: October 30, 2008.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). This amendment request would revise PPL Susquehanna, LLC, Unit 2 (PPL) Technical Specifications (TSs) Section 2.1.1.2, Minimum Critical Power Ratio Safety Limits (MCPRSLs) for two-loop and single-loop operation.

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

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

Response: No.

The proposed change to the two-loop and single-loop MCPRSLs do not directly or indirectly affect any plant system, equipment, component, or change the processes used to operate the plant. Further, the proposed MCPRSLs were generated using NRC approved methodology and meet the applicable acceptance criteria. Thus, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

Prior to the startup of U2C15, licensing analyses are performed (using NRC approved methodology referenced in TS Section 5.6.5.b) to determine changes in the CPR as a result of anticipated operational occurrences. These results are added to the MCPRSL values to generate the MCPROLs in the COLR [Core Operating Limits Report]. These limits could be different from those specified for the previous Unit 2 COLR. The COLR operating limits thus assure that the MCPRSL will not be exceeded during normal operation or AOOs [anticipated operational occurrences]. Postulated accidents are also analyzed prior to the startup and the results shown to be within the NRC approved criteria.

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

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

Response: No.

The changes to the two-loop and single-loop MCPRSLs do not directly or indirectly affect any plant system, equipment, or component and therefore does not affect the failure modes of any of these items. Thus, the proposed change does not create the possibility of a previously unevaluated operator error or a new single failure.

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

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

Response: No.

Since the proposed changes do not alter any plant system, equipment, component, or processes used to operate the plant, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by TS. The proposed two-loop and single-loop MCPRSLs do not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable TS sections, because the proposed MCPRSLs preserve the required margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Branch Chief: Mark G. Kowal.

Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information (SUNSI) and Safeguards Information (SGI) for Contention Preparation

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

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

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

1. This order contains instructions regarding how potential parties to the proceedings listed above may request access to documents containing sensitive unclassified information (SUNSI and SGI).

2. Within ten (10) days after publication of this notice of opportunity for hearing, any potential party as defined in 10 CFR 2.4 who believes access to SUNSI or SGI is necessary for a response to the notice may request access to SUNSI or SGI. A “potential party” is any person who intends or may intend to participate as a party by demonstrating standing and the filing of an admissible contention under 10 CFR 2.309. Requests submitted later than ten (10) days will not be considered absent a showing of good cause for the late filing, addressing why the request could not have been filed earlier.

3. The requester shall submit a letter requesting permission to access SUNSI and/or SGI 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, MD 20852. The e-mail address for the Office of the Secretary and the Office of the Start Printed Page 4255General Counsel are hearing.docket@nrc.gov and ogcmailcenter.resource@nrc.gov, respectively.[1] The request must include the following information:

a. A description of the licensing action with a citation to this Federal Register notice of opportunity for hearing;

b. 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 (a);

c. If the request is for SUNSI, the identity of the individual requesting access to SUNSI and the requester's need for the information in order to meaningfully participate in this adjudicatory proceeding, particularly why publicly available versions of the application would not be sufficient to provide the basis and specificity for a proffered contention;

d. If the request is for SGI, the identity of the individual requesting access to SGI and the identity of any expert, consultant or assistant who will aid the requester in evaluating the SGI, and information that shows:

(i) Why the information is indispensable to meaningful participation in this licensing proceeding; and

(ii) The technical competence (demonstrable knowledge, skill, experience, training or education) of the requester to understand and use (or evaluate) the requested information to provide the basis and specificity for a proffered contention. The technical competence of a potential party or its counsel may be shown by reliance on a qualified expert, consultant or assistant who demonstrates technical competence as well as trustworthiness and reliability, and who agrees to sign a non-disclosure affidavit and be bound by the terms of a protective order; and

e. If the request is for SGI, Form SF-85, “Questionnaire for Non-Sensitive Positions,” Form FD-258 (fingerprint card), and a credit check release form completed by the individual who seeks access to SGI and each individual who will aid the requester in evaluating the SGI. For security reasons, Form SF-85 can only be submitted electronically, through a restricted-access database. To obtain online access to the form, the requester should contact the NRC's Office of Administration at 301-492-3524.[2] The other completed forms must be signed in original ink, accompanied by a check or money order payable in the amount of $191.00 to the U.S. Nuclear Regulatory Commission for each individual, and mailed to the: Office of Administration, Security Processing Unit, Mail Stop TWB-05-B32M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0012.

These forms will be used to initiate the background check, which includes fingerprinting as part of a criminal history records check. Note: copies of these forms do not need to be included with the request letter to the Office of the Secretary, but the request letter should state that the forms and fees have been submitted as described above.

4. To avoid delays in processing requests for access to SGI, all forms should be reviewed for completeness and accuracy (including legibility) before submitting them to the NRC. Incomplete packages will be returned to the sender and will not be processed.

5. Based on an evaluation of the information submitted under items 2 and 3.a through 3.d, above, the NRC staff will determine within ten days of receipt of the written access 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) there is a legitimate need for access to SUNSI or need to know the SGI requested. For SGI, the need to know determination is made based on whether the information requested is necessary (i.e., indispensable) for the proposed recipient to proffer and litigate a specific contention in this NRC proceeding [3] and whether the proposed recipient has the technical competence (demonstrable knowledge, skill, training, education, or experience) to evaluate and use the specific SGI requested in this proceeding.

6. If standing and need to know SGI are shown, the NRC staff will further determine based upon completion of the background check whether the proposed recipient is trustworthy and reliable. The NRC staff will conduct (as necessary) an inspection to confirm that the recipient's information protection systems are sufficient to protect SGI from inadvertent release or disclosure. Recipients may opt to view SGI at the NRC's facility rather than establish their own SGI protection program to meet SGI protection requirements.

7. A request for access to SUNSI or SGI will be granted if:

a. The request has demonstrated that there is a reasonable basis to believe that a potential party is likely to establish standing to intervene or to otherwise participate as a party in this proceeding;

b. The proposed recipient of the information has demonstrated a need for SUNSI or a need to know for SGI, and that the proposed recipient of SGI is trustworthy and reliable;

c. The proposed recipient of the information has executed a Non-Disclosure Agreement or Affidavit and agrees to be bound by the terms of a Protective Order setting forth terms and conditions to prevent the unauthorized or inadvertent disclosure of SUNSI and/or SGI; and

d. The presiding officer has issued a protective order concerning the information or documents requested.[4] Any protective order issued shall provide that the petitioner must file SUNSI or SGI contentions 25 days after receipt of (or access to) that information. 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 or SGI contentions by that later deadline.

8. If the request for access to SUNSI or SGI is granted, the terms and conditions for access to sensitive unclassified information will be set forth in a draft protective order and affidavit of non-disclosure appended to a joint motion by the NRC staff, any other affected parties to this proceeding,[5] and the petitioner(s). If the diligent efforts by the relevant parties or petitioner(s) fail to result in an agreement on the terms and conditions for a draft protective order or non-disclosure affidavit, the relevant parties Start Printed Page 4256to the proceeding or the petitioner(s) should notify the presiding officer within ten (10) days, describing the obstacles to the agreement.

9. If the request for access to SUNSI is denied by the NRC staff or a request for access to SGI is denied by NRC staff either after a determination on standing and need to know or, later, after a determination on trustworthiness and reliability, the NRC staff shall briefly state the reasons for the denial. Before the Office of Administration makes an adverse determination regarding access, the proposed recipient must be provided an opportunity to correct or explain information. The requester may challenge the NRC staff's adverse determination with respect to access to SUNSI or with respect to standing or need to know for SGI by filing a challenge within ten (10) 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. In the same manner, an SGI requester may challenge an adverse determination on trustworthiness and reliability by filing a challenge within fifteen (15) days of receipt of that determination.

In the same manner, 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 within ten (10) days of the notification by the NRC staff of its grant of such a request.

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.[6]

10. 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/or SGI, 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.

Start Signature

Dated at Rockville, Maryland, this 13th day of January 2009.

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 (SUNSI) and Safeguards Information (SGI) in This Proceeding

DayEvent/activity
0Publication of Federal Register notice/other notice of proposed action and opportunity for hearing, including order with instructions for access requests.
10Deadline for submitting requests for access to SUNSI and/or SGI 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; demonstrating that access should be granted (e.g., showing technical competence for access to SGI); and, for SGI, including application fee for fingerprint/background check.
60Deadline for submitting petition for intervention containing: (i) Demonstration of standing; (ii) all contentions whose formulation does not require access to SUNSI and/or SGI (+25 Answers to petition for intervention; +7 petitioner/requestor reply).
20NRC 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 (1) need for SUNSI or (2) need to know for SGI. (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). If NRC staff makes the finding of need to know for SGI and likelihood of standing, NRC staff begins background check (including fingerprinting for a criminal history records check), information processing (preparation of redactions or review of redacted documents), and readiness inspections.
25If NRC staff finds no “need,” “need to know,” or 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.
190(Receipt +180) If NRC staff finds standing, need to know for SGI, and trustworthiness and reliability, deadline for NRC staff to file motion for Protective Order and draft Non-disclosure Affidavit (or to make a determination that the proposed recipient of SGI is not trustworthy or reliable). Note: Before the Office of Administration makes an adverse determination regarding access, the proposed recipient must be provided an opportunity to correct or explain information.
205Deadline for petitioner to seek reversal of a final adverse NRC staff determination either before the presiding officer or another designated officer.
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 and/or SGI consistent with decision issuing the protective order.
Start Printed Page 4257
A + 28Deadline for submission of contentions whose development depends upon access to SUNSI and/or SGI. 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 or SGI contentions by that later deadline.
A + 53(Contention receipt +25) Answers to contentions whose development depends upon access to SUNSI and/or SGI.
A + 60(Answer receipt +7) Petitioner/Intervenor reply to answers.
BDecision on contention admission.
End Preamble

Footnotes

1.  See footnote 6. 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 and/or SGI under these procedures should be submitted as described in this paragraph.

Back to Citation

2.  The requester will be asked to provide his or her full name, Social Security number, date and place of birth, telephone number, and e-mail address. After providing this information, the requester usually should be able to obtain access to the online form within one business day.

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3.  Broad SGI requests under these procedures are thus highly unlikely to meet the standard for need to know; furthermore, staff redaction of information from requested documents before their release may be appropriate to comport with this requirement. These procedures do not authorize unrestricted disclosure or less scrutiny of a requester's need to know than ordinarily would be applied in connection with an already-admitted contention.

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4.  If a presiding officer has not yet been designated, the Chief Administrative Judge will issue such orders, or will appoint a presiding officer to do so.

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5.  Parties/persons other than the requester and the NRC staff will be notified by the NRC staff of a favorable access determination (and may participate in the development of such a motion and protective order) if it concerns SUNSI and if the party/person's interest independent of the proceeding would be harmed by the release of the information (e.g., as with proprietary information).

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6.  As of October 15, 2007, the NRC's final “E-Filing Rule” became effective. See Use of Electronic Submissions in Agency Hearings (72 FR 49139; Aug. 28, 2007). Requesters should note that the filing requirements of that rule apply to appeals of NRS staff determinations (because they must be served on a presiding officer or the Commission, as applicable), but not to the initial SUNSI/SGI requests submitted to the NRC staff under these procedures.

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[FR Doc. E9-1152 Filed 1-22-09; 8:45 am]

BILLING CODE 7590-01-P