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

AGENCY:

Nuclear Regulatory Commission.

ACTION:

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

SUMMARY:

The U.S. Nuclear Regulatory Commission (NRC) received and is considering approval of five amendment requests. The amendment requests are for Palisades Nuclear Plant (PNP); Donald C. Cook Nuclear Plant, Units 1 and 2; Fort Calhoun Station, Unit No. 1; Diablo Canyon Nuclear Power Plant, Units 1 and 2; and Hope Creek Generating Station. For each amendment request, the NRC proposes to determine that they involve no significant hazards consideration. In addition, each amendment request contains sensitive unclassified non-safeguards information (SUNSI).

DATES:

Comments must be filed by July 7, 2016. A request for a hearing must be filed by August 8, 2016. Any potential party as defined in § 2.4 of title 10 of the Code of Federal Regulations (10 CFR), who believes access to SUNSI is necessary to respond to this notice must request document access by June 17, 2016.

ADDRESSES:

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

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0096. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1384, email: Janet.Burkhardt@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2016-0096 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

• Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0096.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2016-0096, facility name, unit number(s), application date, and subject in your comment submission.

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

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

II. Background

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

This notice includes notices of amendments containing SUNSI.

III. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

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

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

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

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one 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 with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with NRC regulations, policies and procedures.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii). 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 unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by August 8, 2016. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions for leave to intervene set forth in this section, except that under § 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c).

If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by August 8, 2016.

B. Electronic Submissions (E-Filing)

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

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

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

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

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

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

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

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

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

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

Date of amendment request: March 3, 2016. A publicly-available version is in ADAMS under Accession No. ML16075A103.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would revise the PNP Technical Specifications (TS), Section 5.5.8, “Steam Generator (SG) Program,” and Section 5.6.8, “Steam Generator Tube Inspection Report.” Specifically, the licensee requested to implement an alternate repair criteria (ARC), that invokes a C-Star inspection length (C*), on a permanent basis for the cold-leg side of the SGs' tubesheet.

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.

Previously evaluated accidents are initiated by the failure of plant structures, systems, or components. The proposed change alters the SG cold leg repair criteria by limiting tube inspections length in the cold leg tubesheet, to the safety significant section, C* length, and, as such, does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. Therefore, the proposed change has no significant effect upon previously evaluated accident probabilities or consequences.

The proposed amendment to revise the PNP SG tube repair criteria in TS 5.5.8c, does not involve a significant increase in the probability of an accident previously evaluated. Alternate repair criteria are being proposed for the cold leg side of the SGs that duplicate the current alternate repair criteria for the hot leg side of the SGs, in TS 5.5.8c.1. The proposed SG tube inspection length maintains the existing design limits of the SGs and therefore does not increase the probability or consequences of an accident involving a tube rupture or primary to secondary accident-induced leakage, as previously evaluated in the PNP Updated Final Safety Analysis Report (UFSAR). Also, the Nuclear Energy Institute (NEI) Steam Generator Program Guidelines (NEI 97-06) [(ADAMS Accession No. ML111310708)] performance criteria for structural integrity and accident-induced leakage, which are incorporated in PNP TS 5.5.8, would continue to be satisfied.

Implementing an alternate repair criteria would allow SG tubes with flaws below the C* length to remain in service. The potential consequences to leaving these flawed tubes inservice are tube burst, tube pullout, and accident induced tube leakage. Tube burst is prevented for a tube with defects within the tubesheet region because of the constraint provided by the tubesheet. Tube pullout could result from the axial forces induced by primary to secondary differential pressures that occur during the bounding event of the main steam line break. A joint industry test program report, WCAP-16208-P, NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions, Revision 1, May 2005 [(Non-proprietary version under ADAMS Accession No. ML051520417)], has defined the non-degraded tube to tubesheet joint length (C*) required to preclude tube pullout and maintain acceptable primary to secondary accident-induced leakage, conservatively assuming a 360 degree circumferential through wall crack exists immediately below this C* length.

The PNP UFSAR Sections 14.14, Steam Line Rupture Incident, 14.15, Steam Generator Tube Rupture with a Loss of Offsite Power, and 14.16, Control Rod Ejection, primary coolant system leakage limit is 0.3 gallon per minute (gpm) (432 gallons per day) in the unaffected SG. For the tube rupture accident, this 0.3 gpm leakage is in addition to the break flow rate associated with the rupture of a single SG tube. The WCAP-16208-P report used a primary to secondary accident-induced leakage criteria value of 0.1 gpm to derive the C* length. Use of 0.1 gpm ensures that the PNP TS limiting accident-induced leakage of 0.3 gpm is met.

For PNP, the derived C* length for the cold leg side of the SGs is 12.5 inches, which is the same C* length, as the current TS, for the hot leg side of the SGs. Any degradation below the C* length is shown by test results and analysis to meet the NEI 97-06 performance criteria, thereby precluding an increased probability of a tube rupture event or an increase in the consequences of a steam line rupture incident or control rod ejection accident.

Therefore, the C* lengths for the SG hot and cold legs provide assurance that the NEI 97-06 requirements for tube burst and leakage are met and that they conservatively derived maximum combined leakage from both tubesheet joints (hot and cold legs) is less than 0.2 gpm at accident conditions. This combined leakage criterion of 0.2 gpm in the faulted loop retains margin against the PNP TS allowable accident-induced leakage of 0.3 gpm per SG.

In summary, the proposed changes to the PNP TS maintain existing design limits, meet the performance criteria of NEI 97-06 and Regulatory Guide 1.121 [ADAMS Accession No. ML003739366], and the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated in the UFSAR.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment provides for an alternate repair criteria that excludes the lower portion of the steam generator cold leg tubes from inspection below a C* length by implementing an alternate repair criteria. It does not affect the design of the SGs or their method of operation. It does not impact any other plant system or component. Plant operation will not be altered, and all safety functions will continue to perform as previously assumed in the accident analysis.

The proposed amendment does not introduce any new equipment, change existing equipment, create any new failure modes for existing equipment, nor introduce any new malfunctions resulting from tube degradation. SG tube integrity is shown to be maintained for all plant conditions upon implementation of the proposed alternate repair criteria for the SG cold leg tubesheet region.

The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated because SG tube leakage limits and structural integrity would continue to be maintained during all plant conditions upon implementation of the proposed alternate repair criteria to the PNP TSs. The alternate repair criteria does not introduce any new mechanisms that might result in a different kind of accident from those previously evaluated. Even with the limiting circumstances of a complete circumferential separation (360 degree through wall crack) of a tube below the C* length, tube pullout is precluded and leakage is predicted to be maintained with the TS and accident analysis limits during all plant conditions.

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

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

Response: No.

The proposed change provides an alternate repair criteria for the SG cold leg that invokes a C* inspection length criteria. The proposed amendment does not involve a significant reduction in a margin of safety since design SG primary to secondary leakage limits have been analyzed to continue to be met. This will ensure that the SG cold legs tubes continue to function as a primary coolant system boundary by maintaining their integrity. Tube integrity includes both structural and leakage integrity. The proposed cold leg tubesheet inspection C* depth, of 12.5 inches below the bottom of the cold-leg expansion transition or top of the cold-leg tubesheet, which is lower, would ensure tube integrity is maintained during normal and accident conditions because any degradation below C* is shown by test results and analyses to be acceptable.

Operation with potential tube degradation below the proposed C* cold leg inspection length within the tubesheet region of the SG tubing meets the recommendation of NEI 97-06 SG program guidelines. Additionally, the proposed changes also maintain the structural and accident-induced leakage integrity as required by NEI 97-06.

The total leakage from an undetected flaw population below the C* inspection length for the cold leg tubesheet under postulated accident conditions is accounted for, in order to assure it is within the bounds of the accident analysis.

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

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

Attorney for licensee: Mrs. Jeanne Cho, Senior Counsel, Entergy Services, Inc., 440 Hamilton Ave., White Plains, New York 10601.

NRC Branch Chief: David J. Wrona.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant (CNP), Units 1 and 2, Berrien County, Michigan

Date of amendment request: March 14, 2016. A publicly-available version is in ADAMS under Accession No. ML16077A029.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would revise the operating license to extend the completion date for full implementation of the CNP Cyber Security Plan (CSP).

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 of occurrence or consequences of an accident previously evaluated?

Response: No.

The amendment proposes a change to the CNP Unit 1 and Unit 2 CSPs Milestone 8 full implementation date as set forth in the CNP CSP Implementation Schedule. The revision of the full implementation date for the CNP CSP does not involve modifications to any safety-related structures, systems or components (SSCs). Rather, the implementation schedule provides a timetable for fully implementing the CNP CSP. The CSP describes how the requirements of 10 CFR 73.54 are to be implemented to identify, evaluate, and mitigate cyber attacks up to and including the design basis cyber attack threat, thereby achieving high assurance that the facility's digital computer and communications systems and networks are adequately protected from cyber attacks. The revision of the CNP CSP Implementation Schedule will not alter previously evaluated design basis accident analysis assumptions, add any accident initiators, modify the function of the plant safety-related SSCs, or affect how any plant safety-related SSCs are operated, maintained, modified, tested, or inspected.

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

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

Response: No.

A revision to the CSP Implementation Schedule does not require any plant modifications. The proposed revision to the CSP Implementation Schedule does not alter the plant configuration, require new plant equipment to be installed, alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected. Revision of the CNP CSP Implementation Schedule does not introduce new equipment that could create a new or different kind of accident, and no new equipment failure modes are created. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed amendment.

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

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

Response: No.

Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed amendment does not alter the way any safety-related SSC functions and does not alter the way the plant is operated. The CSP, as implemented by milestones 1-7, provides assurance that safety-related SSCs are protected from cyber attacks. The proposed amendment does not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. The proposed amendment has no effect on the structural integrity of the fuel cladding, reactor coolant pressure boundary, or containment structure.

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: Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, Michigan 49106.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: April 4, 2016. A publicly-available version is in ADAMS under Accession No. ML16103A348.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendment would modify License Condition D, Fire Protection Program. License Amendment No. 275, issued June 16, 2014 (ADAMS Accession No. ML14098A092), implemented the licensee's transition to a risk-informed, performance-based fire protection program based on National Fire Protection Association Standard (NFPA) 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition.” As part of the Transition License Conditions included in Amendment No. 275, the licensee committed to implement certain plant modifications as stated in Paragraph 3.D.(3)(b) of Renewed Facility Operating License No. DPR-40. Based on updated fire modeling assumptions, the licensee is proposing to withdraw the commitments in REC-119 and REC-120 due to the fact that they are not necessary to meet the performance requirements of the risk-informed fire protection standard.

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 Updated Safety Analysis Report (USAR) documents the analyses of design basis accidents (DBA) at FCS. The proposed amendment does not adversely affect accident initiators nor alter design assumptions, conditions, or configurations of the facility and does not adversely affect the ability of structures, systems, or components (SSCs) to perform their design functions. SSCs required to safely shutdown the reactor and to maintain it in a safe shutdown condition will remain capable of performing their design functions.

The proposed amendment makes no physical changes to the plant and does not change the manner in which plant systems are controlled. Therefore, the implementation of the proposed amendment does not increase the probability of any accident previously evaluated.

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

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

Response: No.

Operation of FCS in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Any scenario or previously analyzed accident with off-site dose was included in the evaluation of DBAs documented in the USAR. The proposed change does not alter the requirements or function for systems required during accident conditions. Implementation of the proposed amendment will not change the previous conclusion that the fire protection licensing basis which complies with the requirements of 10 CFR 50.48(a) and (c) and the guidance in [Regulatory Guide (RG)] 1.205, Revision 0 [Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, May 2006, available under ADAMS Accession No. ML061100174], will not result in new or different accidents.

The proposed amendment does not adversely affect accident initiators nor alter design assumptions, conditions, or configurations of the facility. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shutdown the reactor and maintain it in a safe shutdown condition remain capable of performing their design functions.

The purpose of the proposed amendment is to modify a commitment made as a licensing condition under Amendment No. 275 which implemented OPPD's transition to NFPA 805. The proposed amendment is not intended to reduce or, in any way, adversely affect compliance with NFPA 805 and is supported by engineering analyses that continue to demonstrate compliance with 10 CFR 50.48(a) and (c) and the guidance in RG 1.205, Revision 0.

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

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

Response: No.

Operation of FCS in accordance with the proposed amendment does not involve a significant reduction in the margin of safety. The proposed amendment does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed to mitigate accidents in the USAR. This amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shutdown the reactor and to maintain it in a safe shutdown condition remain capable of performing their design functions.

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

Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: October 26, 2011, as supplemented by letters dated December 20, 2011; April 2, April 30, June 6, August 2, September 11, November 27, and December 5, 2012; March 7, March 25, April 30, May 9, May 30, and September 17, 2013; April 24 and April 30, 2014; February 2 and June 22, 2015; and January 25 and February 11, 2016. Publicly-available versions are in ADAMS under Accession Nos. ML113070457, ML113610541, ML12094A072, ML12131A513, ML121700592, ML122220135, ML12256A308, ML130040687, ML12342A149, ML13267A127, ML130930344, ML13121A089, ML13130A059, ML131540159, ML13261A354, ML14205A031, ML14121A002, ML15062A386, ML15173A469, ML16049A006, and ML16061A481, respectively.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendments would revise the facility operating licenses to allow the permanent replacement of the current 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 (a Westinghouse Electric Company), Advanced Logic System. The amendments would also incorporate a revised definition of Channel Operational Test in Technical Specification (TS) 1.1, “Definitions.”

The license amendment request was originally noticed in the Federal Register on June 5, 2012 (77 FR 33243). The notice is being reissued in its entirety to include a revised description of the amendment request (change to TS 1.1).

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

1. 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 [n]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 requests involve no significant hazards consideration.

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: September 21, 2015, as supplemented by letter dated November 19, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML15265A223 and ML15323A268, respectively.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed amendment would allow for the replacement and upgrade of the existing analog Average Power Range Monitor (APRM) sub-system of the Neutron Monitoring System with General Electric-Hitachi digital Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) system. The PRNM upgrade also includes Oscillation Power Range Monitor (OPRM) capability and will allow full APRM, Rod Block Monitor (RBM), Technical Specification Improvement Program implementation, and will include application of Technical Specification Task Force Traveler-493, “Clarify Application of Setpoint Methodology for LSSS [Limiting Safety System Setting] Functions,” to affected PRNM functions.

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 accidents occurring is not affected by the PRNM system, as the PRNM system is not the initiator of any accident and does not interact with equipment whose failure could cause an accident. The transition from flow-biased to power-biased RBM does not increase the probability of an accident; the RBM is not involved in the initiation of any accident. The regulatory criteria established for the APRM, OPRM, and RBM systems will be maintained with the installation of the upgraded PRNM system. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The consequences of accidents are not affected by the PRNM system, as the setpoints in the PRNM system will be established so that all analytical limits are met. The unavailability of the new system will be equal to or less than the existing system and, as a result, the scram reliability will be equal to or better than the existing system. No new challenges to safety-related equipment will result from the PRNM system modification. The change to power biased RBM allows for Rod Withdrawal Error (RWE) analyses performed for each future reload to take credit for rod blocks during the rod withdrawal transients. The results of the RWE event analysis will be used in establishing the cycle specific operating limits for the fuel. The proposed change will also replace the currently installed and NRC approved Asea Brown Boveri (ABB) OPRM Option III long-term stability solution with an NRC approved General Electric-Hitachi (GEH) Detect and Suppress Solution—Confirmation Density (DSS-CD) stability solution (reviewed and approved by the NRC in Reference 2, Licensing Topical Report). The OPRM meets the GDC [General Design Criteria] 10, “Reactor Design,” and 12, “Suppression of Reactor Power Oscillations,” requirements by automatically detecting and suppressing design basis thermal hydraulic oscillations to protect specified fuel design limits. Therefore, the proposed change does not involve a significant increase in the consequences of an 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 amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The components of the PRNM system will be supplied to equivalent or better design and qualification criteria than is currently required for the plant. Equipment that could be affected by [the] PRNM system has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or system interaction mode was identified. Therefore, the upgraded PRNM system will not adversely affect plant equipment.

The new PRNM system uses digital equipment that has software controlled digital processing points and software controlled digital processing compared to the existing PRNM system that uses mostly analog and discrete component processing (excluding the existing OPRM). Specific failures of hardware and potential software common cause failures are different from the existing system. The effects of potential software common cause failure are mitigated by specific hardware design and system architecture as discussed in Section 6.0 of the NUMAC PRNM LTR [Licensing Topical Report], and supported by a plant specific evaluation. The transition from a flow-biased RBM to a power dependent RBM does not change its function to provide a control rod block when specified setpoints are reached. The change does not introduce a sequence of events or introduce a new failure mode that would create a new or different type of accident. Failure(s) of the system have the same overall effect as the present design. No new or different kind of accident is introduced. Therefore, the PRNM system will not adversely affect plant equipment.

The currently installed APRM System is replaced with a NUMAC PRNM system that performs the existing power range monitoring functions and adds an OPRM to react automatically to potential reactor thermal-hydraulic instabilities.

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

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

Response: No.

The proposed TS changes associated with the NUMAC PRNM system implement the constraints of the NUMAC PRNM system design and related stability analyses. The NUMAC PRNM system change does not impact reactor operating parameters or the functional requirements of the PRNM system. The replacement equipment continues to provide information, enforce control rod blocks, and initiate reactor scrams under appropriate specified conditions. The power dependent RBM will continue to prevent rod withdrawal when the power-dependent RBM rod block setpoint is reached. The MCPR [Minimum Critical Power Ratio] and Linear Heat Generation Rate (LHGR) thermal limits will be developed on a cycle specific basis to ensure that fuel thermal mechanical design bases remain within the licensing limits during a control rod withdrawal error event and to ensure that the MCPR SL [Safety Limit] will not be violated as a result of a control rod withdrawal error event.

The proposed change does not reduce safety margins. The replacement PRNM equipment has improved channel trip accuracy compared to the current analog system, and meets or exceeds system requirements previously assumed in setpoint analysis. The power dependent RBM will support cycle specific RWE analysis ensuring fuel limits are not exceeded. Thus, the ability of the new equipment to enforce compliance with margins of safety equals or exceeds the ability of the equipment which it replaces.

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

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

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

NRC Branch Chief: Douglas A. Broaddus.

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

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

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

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

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

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

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

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

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

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

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

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

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

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

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

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

G. Review of Denials of Access.

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

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

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

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

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.

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