[Federal Register Volume 77, Number 108 (Tuesday, June 5, 2012)]
[Notices]
[Pages 33243-33252]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2012-13426]


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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0125]


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.

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

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and are publicly available, 
by searching on http://www.regulations.gov under Docket ID NRC-2012-
0125. You may submit comments by the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0125. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

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

[[Page 33244]]

B. Submitting Comments

    Please include Docket ID NRC-2012-0125 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information in comment submissions that you do not want to be publicly 
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS, and the NRC does not edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information in their comment submissions 
that they do not want to be publicly disclosed. Your request should 
state that the NRC will not edit comment submissions to remove such 
information before making the comment submissions available to the 
public or entering the comment submissions into ADAMS.

II. Background

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

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Rules of 
Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. The NRC's regulations are accessible electronically from the NRC 
Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed within 60 days, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to

[[Page 33245]]

participate fully in the conduct of the hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. With 
respect to copyrighted works, except for limited excerpts that serve 
the purpose of the adjudicatory filings and would

[[Page 33246]]

constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).

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

    Date of amendment request: February 28, 2012. Publicly available 
versions of the amendment request and its attachment are available in 
ADAMS under Accession Nos. ML12061A288 and ML12061A289.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would revise the Palisades Nuclear Plant Technical 
Specifications (TS) to support the replacement of the Region I spent 
fuel pool (SFP) storage racks with new neutron absorber Metamic-
equipped racks. Degradation of the present neutron absorber, 
Carborundum[supreg], has resulted in reduced SFP storage capacity. The 
replacement of the SFP storage racks will allow recovery of the 
currently unusable storage locations in the SFP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The probability of any accident previously evaluated is not 
significantly increased by the proposed changes to the Region I 
spent fuel pool (SFP) storage racks. The probabilities of an 
accidental fuel assembly drop or misloading is primarily influenced 
by the methods used to lift and move these loads. The method of 
handling fuel is not changed since the same equipment and procedures 
will be used. Work in the SFP area will be controlled and performed 
in accordance with written procedures. Any movement of fuel 
assemblies required to be performed to support the modification will 
be performed in the same manner as during normal operations. 
Replacing the Region I SFP storage racks does not have a significant 
impact on the frequency of occurrence for any accident previously 
evaluated. Additionally, the probabilities of a seismic event, boron 
dilution, or loss of SFP cooling flow are not influenced by the 
proposed changes. Therefore, the proposed change will not involve a 
significant increase in the probability of occurrence of any event 
previously analyzed.
    TS 3.7.15, SFP Boron Concentration, requires a minimum boron 
concentration of 1720 ppm, which bounds the analysis for the 
proposed amendment. Soluble boron is maintained in the SFP water as 
required by the TS and controlled by procedures. The criticality 
safety analyses for Region I and for Region II of the SFP credit the 
same soluble boron concentration of 850 ppm to maintain a Keff <= 
0.95 under normal conditions and 1350 ppm to maintain a Keff <= 0.95 
under accident scenarios as does the analysis for the proposed 
change for Region I. In crediting soluble boron, in Region 1, the 
SFP criticality analysis would have no effect on normal pool 
operation and maintenance. Thus, there is no change to the 
probability or the consequences of the boron dilution event in the 
SFP.
    The consequences of the dropped spent fuel assembly in the SFP 
have been re-evaluated for the proposed change by analyzing a 
potential impact on the replacement racks. The results show that the 
postulated accident of a fuel assembly striking the top of the 
replacement racks would not distort the racks sufficiently to impair 
their functionality. The minimum subcriticality margin (i.e., 
neutron multiplication factor (Keff) less than or equal to 0.95) 
will be maintained. The structural damage to the fuel building, pool 
liner, and fuel assembly resulting from a dropped fuel assembly 
striking the pool floor or another assembly located in the racks is 
primarily dependent on the mass of the falling object and drop 
height. Since these two parameters are not changed by the proposed 
modification, the postulated structural damage to these items 
remains unchanged. The radiological dose at the exclusion area 
boundary has been evaluated and found to remain well below levels 
established by regulatory guidance.
    The consequences of a loss of SFP cooling were evaluated and 
found to not involve a significant increase as a result of the 
proposed changes. The concern with this accident is a reduction of 
SFP water inventory from bulk pool boiling resulting in uncovering 
fuel assemblies. This situation could lead to fuel failure and 
subsequent significant increase in offsite dose. Loss of SFP cooling 
is mitigated by ensuring that a sufficient time lapse exists between 
the loss of forced cooling and the uncovering of fuel. This period 
of time is compared against a reasonable period to re-establish 
cooling or supply an alternative water source. Evaluation of this 
accident includes determination of the time to boil. This time 
period is much less than the onset of any significant increase in 
offsite dose, since once boiling begins it would have to continue 
unchecked until the pool water surface was lowered to the point of 
exposing active fuel. The time to boil represents the onset of loss 
of pool water inventory and is used as a gauge for establishing the 
comparison of consequences before and after a rack replacement 
project. The heatup rate in the SFP is a nearly linear function of 
the fuel decay heat load. The thermal-hydraulic analysis determined 
that the minimum time to boil is at least 1.8 hours subsequent to 
complete loss of forced cooling. In the unlikely event that all pool 
cooling is lost, sufficient time will still be available subsequent 
to the proposed changes for the operators to provide alternate means 
of cooling before the water shielding above the top of the racks 
falls below an acceptable level.
    The consequences of a design basis seismic event are not 
increased. The consequences of this event are evaluated on the basis 
of subsequent fuel damage or compromise of the fuel storage or 
building configurations leading to radiological or criticality 
concerns. The replacement racks have been analyzed and were found to 
be safe during seismic motion. Fuel has been determined to remain 
intact and the storage racks maintain the fuel and fixed neutron 
absorber configurations subsequent to a seismic event. The 
replacement racks do not impact the pool walls. The structural 
capability of the pool and liner will not be exceeded under the 
appropriate combinations of dead weight, thermal, and seismic loads. 
The fuel building structure will remain intact during a seismic 
event and will continue to adequately support and protect the fuel 
racks, storage array, and pool moderator and coolant.
    The consequence of a fuel misloading accident has been analyzed 
for the worst possible storage configuration subsequent to the 
proposed modification and it has been shown that the consequences 
remain acceptable with respect to the same criteria used previously.
    In summary, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The existing TS allow storage of fuel assemblies with a maximum 
planar average U-235 enrichment of 4.54 weight percent. The proposed 
change, for the replacement Region I fuel storage racks, would 
increase the maximum planar average U-235 enrichment to 4.95 weight 
percent. Fuel would be allowed in all the storage cell locations in 
the Metamic \TM\ equipped Region I storage racks. Therefore, the 
possibility of misplacing a fuel assembly in the replacement fuel 
storage racks, with greater enrichment than allowed in certain 
storage locations in Region I, for the Carborundum equipped fuel 
storage racks would be eliminated, for the replacement Metamic \TM\ 
equipped fuel storage racks. Changing the enrichment and allowing 
fuel storage in all the storage locations in the Metamic \TM\ 
equipped Region I storage racks does not create a new or different 
kind of accident from any previously evaluated.
    No new or different activities are introduced in the replacement 
of the fuel storage racks other than the physical removal of the 
existing racks and the installation of

[[Page 33247]]

the new Metamic \TM\ equipped fuel storage racks. An accident of a 
rack dropping onto stored spent fuel or the pool floor liner is not 
a postulated event due to the defense-in-depth approach to be taken. 
A rack lifting rig will be introduced to remove the existing Region 
I racks and to install the replacement racks. The temporary lift 
items are designed to meet the requirements of NUREG-0612, Control 
of Heavy Loads, and ANSI N14.6, Radioactive Materials--Special 
Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 
kg) or More. The lift rig and rack would be lifted using the fuel 
building crane main hook. This crane and main hook satisfy the 
single failure proof criteria of NUREG-0554, Single Failure Proof 
Cranes for Nuclear Power Plants. A rack drop event is a ``heavy load 
drop'' over the SFP. A lifted rack will not be allowed to travel 
over any stored fuel assemblies, thus a rack drop onto fuel is 
precluded. A rack drop to the pool liner is not a postulated event. 
All movements of heavy loads over the pool will comply with the 
applicable administrative controls and guidelines. Therefore, the 
activities for removal and installation of the fuel storage racks 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change does not alter the operating requirements of 
the plant or of the equipment credited in the mitigation of the 
design basis accidents. The changes would not affect the SFP cooling 
system or the SFP makeup systems. The proposed change does not 
affect any of the important parameters required to ensure safe fuel 
storage.
    In summary, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The function of the SFP is to store the fuel assemblies in a 
subcritical and cool-able configuration through all environmental 
and abnormal loadings, such as an earthquake or fuel assembly drop. 
The replacement Region I fuel storage rack design must meet all 
applicable requirements for safe storage and be functionally 
compatible with the SFP.
    Detailed analysis has shown, with a 95 percent probability at a 
95 percent confidence level, that the Keff of the Region I fuel 
storage racks in the SFP, including uncertainties, is less than 1.0 
with unborated water and is less than or equal to 0.95 with 850 ppm 
of soluble boron in the SFP. In addition, the effects of abnormal 
and accident conditions have been evaluated to demonstrate that 
under credible conditions the Keff will not exceed 0.95 with 1350 
ppm soluble boron credited. The current TS requirement for minimum 
SFP boron concentration is 1720 ppm, which provides assurance that 
the SFP would remain subcritical under normal, abnormal, or accident 
conditions. The margin of safety for subcriticality is maintained by 
having Keff equal to or less than 0.95 under all normal storage, 
fuel handling, and accident conditions, including uncertainties.
    The current analysis basis for the Region I and Region II fuel 
storage racks is a maximum Keff of less than 1.0 when flooded with 
unborated water, and less than or equal to 0.95 when flooded with 
water having a boron concentration of 850 ppm. In addition, the Keff 
in accident or abnormal operating conditions is less than 0.95 with 
1350 ppm of soluble boron. These values are not affected by the 
proposed change. Therefore, the margin of safety is not reduced.
    The mechanical, material, and structural designs of the 
replacement racks have been reviewed in accordance with the 
applicable NRC guidance. The rack materials used are compatible with 
the spent fuel assemblies and the SFP environment. The design of the 
replacement racks preserves the margin of safety during abnormal 
loads such as a dropped fuel assembly. It has been shown that such 
loads will not invalidate the mechanical design and material 
selection to safely store fuel in a cool-able and subcritical 
configuration.
    The thermal-hydraulic and cooling evaluation of the pool 
demonstrated that the pool can be maintained below the specified 
thermal limits under the conditions of the maximum heat load and 
during all credible accident sequences and seismic events. The pool 
temperature will not exceed 150[deg]F during the worst single 
failure of a cooling pump. The maximum local water and fuel cladding 
temperatures in the hot channel will remain below the boiling point. 
The fuel will not undergo any significant heat up after an 
accidental drop of a fuel assembly on top of the rack blocking the 
flow path. A loss of cooling to the pool will allow sufficient time 
(nearly 2 hours) for the operators to intervene and line up 
alternate cooling paths and the means of inventory make-up before 
boiling begins. The thermal limits specified for the evaluations 
performed to support the proposed change are the same as those that 
were used in the previous evaluations.
    Therefore, the proposed change for the replacement of the Region 
I SFP storage racks does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Acting Branch Chief: Istvan Frankl.

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

    Date of application for amendments: November 3, 2011. A publicly 
available version is available in ADAMS under Accession No. 
ML113081441.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would modify the Technical Specifications to include the use 
of neutron absorbing spent fuel pool rack inserts for the purpose of 
criticality control in the spent fuel pools.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Technical Specification (TS) 4.3.1 
to permit installation of NETCO-SNAP-IN[supreg] rack inserts in 
spent fuel storage rack cells. The change is necessary to ensure 
that, with continued Boraflex degradation over time, the effective 
neutron multiplication factor, Keff, is less than or 
equal to 0.95, if the spent fuel pool is fully flooded with 
unborated water as required by 10 CFR 50.68. Because the proposed 
change pertains only to the spent fuel pool, only those accidents 
that are related to movement and storage of fuel assemblies in the 
spent fuel pool could potentially be affected by the proposed 
change.
    The installation of NETCO-SNAP-IN[supreg] rack inserts does not 
result in a significant increase in the probability of an accident 
previously analyzed because there are no changes in the manner in 
which spent fuel is handled, moved, or stored in the rack cells. The 
probability that a fuel assembly would be dropped is unchanged by 
the installation of the NETCO-SNAP-IN[supreg] rack inserts. These 
events involve failures of administrative controls, human 
performance, and equipment failures that are unaffected by the 
presence or absence of Boraflex and the rack inserts.
    The installation of NETCO-SNAP-IN[supreg] rack inserts does not 
result in a significant increase in the consequences of an accident 
previously analyzed because there is no change to the fuel 
assemblies that provide the source term used in calculating the 
radiological consequences of a fuel handling accident. In addition, 
consistent with the current design, only one fuel assembly will be 
moved at a time. Thus, the consequences of dropping a fuel assembly 
onto any other fuel assembly or other structure remain bounded by 
the previously analyzed fuel handling accident. The proposed change 
does not affect the effectiveness of the other engineered design 
features to limit the offsite dose consequences of the limiting fuel 
handling accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

[[Page 33248]]

    Response: No.
    Onsite storage of spent fuel assemblies in the Peach Bottom 
Atomic Power Station (PBAPS), Units 2 and 3 spent fuel pools is a 
normal activity for which PBAPS has been designed and licensed. As 
part of assuring that this normal activity can be performed without 
endangering public health and safety, the ability to safely 
accommodate different possible accidents in the spent fuel pool, 
such as dropping a fuel assembly or misloading a fuel assembly, have 
been analyzed. The proposed spent fuel storage configuration does 
not change the methods of fuel movement or spent fuel storage. The 
proposed change allows for continued use of spent fuel pool storage 
rack cells with degraded Boraflex within those spent fuel pool 
storage rack cells.
    The rack inserts are passive devices. These devices, when inside 
a spent fuel storage rack cell, perform the same function as the 
previously licensed Boraflex neutron absorber panels in that cell. 
These devices do not add any limiting structural loads or affect the 
removal of decay heat from the assemblies. No change in total heat 
load in the spent fuel pool is being made. The devices will maintain 
their design function over the life of the spent fuel pool. The 
existing fuel handling accident, which assumes the drop of a fuel 
assembly, bounds the drop of a rack insert and/or rack insert 
installation tool. This proposed change does not create the 
possibility of misloading an assembly into a spent fuel storage rack 
cell.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    PBAPS TS 4.3.1.1 requires the spent fuel storage racks to 
maintain the effective neutron multiplication factor, 
Keff, less than or equal to 0.95 when fully flooded with 
unborated water, which includes an allowance for uncertainties. 
Therefore, for criticality, the required safety margin is 5% 
including a conservative margin to account for engineering and 
manufacturing uncertainties.
    The proposed change provides a method to ensure that 
Keff continues to be less than or equal to 0.95, thus 
preserving the required safety margin of 5%. The criticality 
analyses demonstrate that the required margin to criticality of 5%, 
including a conservative margin to account for engineering and 
manufacturing uncertainties, is maintained assuming an infinite 
array of fuel with all fuel at the peak reactivity. In addition, the 
radiological consequences of a dropped fuel assembly are unchanged 
because the event involving a dropped fuel assembly onto a spent 
fuel storage rack cell containing a fuel assembly with a rack insert 
is bounded by the radiological consequences of a dropped fuel 
assembly without a rack insert. The proposed change also maintains 
the capacity of the Unit 2 and Unit 3 spent fuel pools to be no more 
than 3,819 fuel assemblies each.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for Licensee: Mr. J. Bradley Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Meena K. Khanna.

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

    Date of amendment request: April 10, 2012. A publically available 
version is available in ADAMS under Accession No. ML12121A527.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
changes would revise the Seabrook Station Technical Specifications 
(TSs). The proposed change would revise TS 6.7.6.k, Steam Generator 
(SG) Program, to exclude a portion of the tubes below the top of the SG 
tube sheet from periodic tube inspections and plugging. The proposed 
change also establishes permanent reporting requirements in TS 6.8.1.7, 
Steam Generator Tube Inspection Report, that were previously 
implemented on a temporary basis.
    Basis for proposed no significant hazards consideration (NSHC) 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of NSHC, which is presented below 
with NRC edits in brackets:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Response: No.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the steam generator (SG) inspection and reporting 
criteria does not have a detrimental impact on the integrity of any 
plant structure, system, or component that initiates an analyzed 
event. The proposed change will not alter the operation of, or 
otherwise increase the failure probability of any plant equipment 
that initiates an analyzed accident.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed change to the SG tube 
inspection and repair criteria are the steam generator tube rupture 
(SGTR) event, the steam line break (SLB), and the feed line break 
(FLB) postulated accidents.
    Addressing the SGTR event, the required structural integrity 
margins of the SG tubes and the tube-to-tubesheet joint over the H* 
distance will be maintained. Tube rupture in tubes with cracks 
within the tubesheet is precluded by the constraint provided by the 
presence of the tubesheet and the tube-to-tubesheet joint. Tube 
burst cannot occur within the thickness of the tubesheet. The tube-
to-tubesheet joint constraint results from the hydraulic expansion 
process, thermal expansion mismatch between the tube and tubesheet, 
and from the differential pressure between the primary and secondary 
side, and tubesheet rotation. The structural margins against burst, 
as discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging 
Degraded PWR [Pressurized-Water Reactor] Steam Generator Tubes,'' 
and Technical Specification (TS) 6.7.6.k, are maintained for both 
normal and postulated accident conditions.
    The proposed change has no impact on the structural or leakage 
integrity of the portion of the tube outside of the tubesheet. The 
proposed change maintains structural and leakage integrity of the SG 
tubes consistent with the performance criteria of TS 6.7.6.k. 
Therefore, the proposed change results in no significant increase in 
the probability of the occurrence of a SGTR accident.
    At normal operating pressures, leakage from tube degradation 
below the proposed limited inspection depth is limited by the tube-
to-tubesheet crevice. Consequently, negligible normal operating 
leakage is expected from degradation below the inspected depth 
within the tubesheet region. The consequences of an SGTR event are 
not affected by the primary-to-secondary leakage flow during the 
event as primary-to-secondary leakage flow through a postulated tube 
that has been pulled out of the tubesheet is essentially equivalent 
to a severed tube. Therefore, the proposed change does not result in 
a significant increase in the consequences of a SGTR.
    The consequences of a SLB or FLB are also not significantly 
affected by the proposed changes. The leakage analysis shows that 
the primary-to-secondary leakage during a SLB/FLB event would be 
less than or equal to that assumed in the Updated Safety Analysis 
Report.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., a SLB/FLB) is 
limited by flow restrictions. These restrictions result from the 
crack and tube-to-tubesheet contact pressures that provide a 
restricted leakage path above the indications and also limit the 
degree of potential crack face opening as compared to free span 
indications.
    The leakage factor of 2.49 for Seabrook Station, for a 
postulated SLB/FLB, has been calculated as shown in References 8, 9 
and 10. For the Condition Monitoring assessment, the component of 
leakage from the prior cycle from below the H* distance will be 
multiplied by a factor of 2.49 and added to the total leakage from 
any other source and compared to the allowable accident induced 
leakage limit. For the Operational Assessment, the difference in the 
leakage between the allowable leakage and the accident induced 
leakage from sources other than the tubesheet expansion region will 
be divided by 2.49 and compared to the observed operational leakage.

[[Page 33249]]

    The probability of a SLB/FLB is unaffected by the potential 
failure of a SG tube as the failure of the tube is not an initiator 
for a SLB/FLB event. SLB/FLB leakage is limited by flow restrictions 
resulting from the leakage path above potential cracks through the 
tube-to-tubesheet crevice. The leak rate during all postulated 
accident conditions that model primary-to-secondary leakage 
(including locked rotor and control rod ejection) has been shown to 
remain within the accident analysis assumptions for all axial and or 
circumferentially orientated cracks occurring 15.21 inches below the 
top of the tubesheet. The assumed accident induced leak rate for 
Seabrook is 500 gallons per day (gpd) during a postulated steam line 
break in the faulted loop. Using the limiting leak rate factor of 
2.49, this corresponds to an acceptable level of operational leakage 
of 200 gpd. Therefore, the TS leak rate limit of 150 gpd provides 
significant added margin against the 500 gpd accident analysis leak 
rate assumption.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Response: No.
    The proposed change that alters the SG inspection and reporting 
criteria does not introduce any new equipment, create new failure 
modes for existing equipment, or create any new limiting single 
failures. Plant operation will not be altered, and all safety 
functions will continue to perform as previously assumed in accident 
analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    Response: No.
    The proposed change that alters the SG inspection and reporting 
criteria maintains the required structural margins of the SG tubes 
for both normal and accident conditions. Nuclear Energy Institute 
97-06, Rev. 3 ``Steam Generator Program Guidelines,'' and NRC 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam 
Generator Tubes,'' are used as the bases in the development of the 
limited hot leg tubesheet inspection depth methodology for 
determining that SG tube integrity considerations are maintained 
within acceptable limits. RG 1.121 describes a method acceptable to 
the NRC for meeting General Design Criteria (GDC) 14, ``Reactor 
Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant System 
Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant Pressure 
Boundary,'' and GDC 32, ``Inspection of Reactor Coolant Pressure 
Boundary,'' by reducing the probability and consequences of a SGTR. 
RG 1.121 concludes that by determining the limiting safe conditions 
for tube wall degradation, the probability and consequences of a 
SGTR are reduced. This RG uses safety factors on loads for tube 
burst that are consistent with the requirements of Section III of 
the American Society of Mechanical Engineers (ASME) Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse WCAP-17071-P 
defines a length of degradation-free expanded tubing that provides 
the necessary resistance to tube pullout due to the pressure induced 
forces, with applicable safety factors applied. Application of the 
limited hot and cold leg tubesheet inspection criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage as described in 
WCAP-17071-P provides significant margin between the accident-
induced leakage assumption and the technical specification leakage 
limit during normal operating conditions when the proposed limited 
tubesheet inspection depth criteria is implemented.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves NSHC.
    Attorney for licensee: M.S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Meena Khanna.

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

    Date of amendment request: October 26, 2011. A publicly available 
version is available in ADAMS under Accession No. ML113070457.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The 
amendment would revise the facility operating licenses to allow the 
permanent replacement of the current Diablo Canyon Power Plant, Units 1 
and 2 (DCPP) Eagle 21 digital process protection system (PPS) with a 
new digital PPS that is based on the Invensys Operations Management 
Tricon Programmable Logic Controller (PLC), Version 10, and the CS 
Innovations, LLC (CSI, a Westinghouse Electric Company), Advanced Logic 
System (ALS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or Consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow Pacific Gas and Electric Company 
to permanently replace the Diablo Canyon Power Plant Eagle 21 
digital process protection system with a new digital process 
protection system that is based on the Invensys Operations 
Management Tricon Programmable Logic Controller, Version 10, and the 
CS Innovations Advanced Logic System. The process protection system 
replacement is designed to applicable codes and standards for 
safety-grade protection systems for nuclear power plants and 
incorporates additional redundancy and diversity features and 
therefore, does not result in an increase in the probability of 
inadvertent actuation or probability of failure to initiate a 
protective function. The process protection system replacement does 
not introduce any new credible failure mechanisms or malfunctions 
that cause an accident. The process protection system replacement 
design will continue to perform the reactor trip system and 
engineered safety features actuation system functions assumed in the 
Final Safety Analysis Report within the response time assumed in the 
Final Safety Analysis Report Chapter 6 and 15 accident analyses.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed change is to permanently replace the current Diablo 
Canyon Power Plant Eagle 21 digital process protection system with a 
new digital process protection system. The process protection system 
performs the process protection functions for the reactor protection 
system that monitors selected plant parameters and initiates 
protective action as required. Accidents that may occur due to 
inadvertent actuation of the process protection system, such as an 
inadvertent safety injection actuation, are considered in the Final 
Safety Analysis Report accident analyses.
    The protection system is designed with redundancy such that a 
single failure to generate an initiation signal in the process 
protection system will not cause failure to trip the reactor nor 
failure to actuate the engineered safeguard features when required. 
Neither will such a single failure cause spurious or inadvertent 
reactor trips or engineered safeguard features actuations because 
coincidence of two or more initiation signals is required for the 
solid state protection system to generate a trip or actuation 
command. If an inadvertent actuation occurs for any reason, existing 
control room alarms and indications will notify the operator to take 
corrective action.
    The process protection system replacement design includes 
enhanced diversity features compared to the current process 
protection system to provide additional assurance that the 
protection system actions credited with

[[Page 33250]]

automatic operation in the Final Safety Analysis Report accident 
analyses will be performed automatically when required should a 
common cause failure occur concurrently with a design basis event.
    The process protection system replacement does not result in any 
new credible failure mechanisms or malfunctions. The current Eagle 
21 process protection system utilizes digital technology and 
therefore the use of digital technology in the process protection 
system replacement does not introduce a new type of failure 
mechanism. Although extremely unlikely, the current Eagle 21 process 
protection system is susceptible to a credible common-cause software 
failure that could adversely affect automatic performance of the 
protection function. The process protection system replacement 
contains new, additional diversity features that prevent a common-
cause software failure from completely disabling the process 
protection system.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The reactor protection system is fundamental to plant safety and 
performs reactor trip system and engineered safety features 
actuation system functions to limit the consequences of Condition II 
(faults of moderate frequency), Condition III (infrequent faults), 
and Condition IV (limiting faults) events. This is accomplished by 
sensing selected plant parameters and determining whether 
predetermined instrument settings are being exceeded. If 
predetermined instrument settings are exceeded, the reactor 
protection system sends actuation signals to trip the reactor and 
actuate those components that mitigate the severity of the accident.
    The process protection system replacement design will continue 
to perform the reactor trip system and engineered safety features 
actuation functions assumed in the Final Safety Analysis Report 
within the response time assumed Final Safety Analysis Report 
Chapter 6 and 15 accident analyses. The use of the process 
protection system replacement does not result in a design basis or 
safety limit being exceeded or changed. The change to the process 
protection system has no impact on the reactor fuel, reactor vessel, 
or containment fission product barriers. The reliability and 
availability of the reactor protection system is improved with the 
process protection system replacement, and the reactor protection 
system will continue to effectively perform its function of sensing 
plant parameters to initiate protective actions to limit or mitigate 
events.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, 77 Beale Street, Room 2496, Mail Code B30A, San 
Francisco, CA 94105.
    NRC Branch Chief: Michael T. Markley.

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

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

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

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

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

    A. This Order contains instructions regarding how potential parties 
to this proceeding may request access to documents containing Sensitive 
Unclassified Non-Safeguards Information (SUNSI).
    B. Within 10 days after publication of this notice of hearing and 
opportunity to petition for leave to intervene, any potential party who 
believes access to SUNSI is necessary to respond to this notice may 
request such access. A ``potential party'' is any person who intends to 
participate as a party by demonstrating standing and filing an 
admissible contention under 10 CFR 2.309. Requests for access to SUNSI 
submitted later than 10 days after publication will not be considered 
absent a showing of good cause for the late filing, addressing why the 
request could not have been filed earlier.
    C. The requestor shall submit a letter requesting permission to 
access SUNSI to the Office of the Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, and provide a copy to the Associate General 
Counsel for Hearings, Enforcement and Administration, Office of the 
General Counsel, Washington, DC 20555-0001. The expedited delivery or 
courier mail address for both offices is: U.S. Nuclear Regulatory 
Commission, 11555 Rockville Pike, Rockville, Maryland 20852. The email 
address for the Office of the Secretary and the Office of the General 
Counsel are Hearing.Docket@nrc.gov and OGCmailcenter@nrc.gov, 
respectively.\1\ The request must include the following information:
---------------------------------------------------------------------------

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

    (1) A description of the licensing action with a citation to this 
Federal Register notice;
    (2) The name and address of the potential party and a description 
of the potential party's particularized interest that could be harmed 
by the action identified in C.(1); and
    (3) The identity of the individual or entity requesting access to 
SUNSI and the requestor's basis for the need for the information in 
order to meaningfully participate in this adjudicatory proceeding. In 
particular, the request must explain why publicly available versions of 
the information requested would not be sufficient to provide the basis 
and specificity for a proffered contention.
    D. Based on an evaluation of the information submitted under 
paragraph C.(3) the NRC staff will determine within 10 days of receipt 
of the request whether:
    (1) There is a reasonable basis to believe the petitioner is likely 
to establish standing to participate in this NRC proceeding; and
    (2) The requestor has established a legitimate need for access to 
SUNSI.
    E. If the NRC staff determines that the requestor satisfies both 
D.(1) and D.(2) above, the NRC staff will notify the requestor in 
writing that access to SUNSI has been granted. The written notification 
will contain instructions on how the requestor may obtain copies of the 
requested documents, and any other conditions that may apply to access 
to those documents. These conditions may include, but are not limited 
to, the signing of a Non-Disclosure Agreement or Affidavit, or 
Protective Order \2\ setting forth terms and conditions to prevent the 
unauthorized or inadvertent disclosure of SUNSI by each individual who 
will be granted access to SUNSI.
---------------------------------------------------------------------------

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

    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

[[Page 33251]]

granted access to that information. However, if more than 25 days 
remain between the date the petitioner is granted access to the 
information and the deadline for filing all other contentions (as 
established in the notice of hearing or opportunity for hearing), the 
petitioner may file its SUNSI contentions by that later deadline.
    G. Review of Denials of Access.
    (1) If the request for access to SUNSI is denied by the NRC staff 
after a determination on standing and need for access, the NRC staff 
shall immediately notify the requestor in writing, briefly stating the 
reason or reasons for the denial.
    (2) The requestor may challenge the NRC staff's adverse 
determination by filing a challenge within 5 days of receipt of that 
determination with: (a) The presiding officer designated in this 
proceeding; (b) if no presiding officer has been appointed, the Chief 
Administrative Judge, or if he or she is unavailable, another 
administrative judge, or an administrative law judge with jurisdiction 
pursuant to 10 CFR 2.318(a); or (c) if another officer has been 
designated to rule on information access issues, with that officer.
    H. Review of Grants of Access. A party other than the requestor may 
challenge an NRC staff determination granting access to SUNSI whose 
release would harm that party's interest independent of the proceeding. 
Such a challenge must be filed with the Chief Administrative Judge 
within 5 days of the notification by the NRC staff of its grant of 
access.
    If challenges to the NRC staff determinations are filed, these 
procedures give way to the normal process for litigating disputes 
concerning access to information. The availability of interlocutory 
review by the Commission of orders ruling on such NRC staff 
determinations (whether granting or denying access) is governed by 10 
CFR 2.311.\3\
---------------------------------------------------------------------------

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

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

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

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.

   Attachment 1--General Target Schedule for Processing and Resolving
Requests for Access to Sensitive Unclassified Non-Safeguards Information
                           in this Proceeding
------------------------------------------------------------------------
           Day                             Event/Activity
------------------------------------------------------------------------
0........................  Publication of Federal Register notice of
                            hearing and opportunity to petition for
                            leave to intervene, including order with
                            instructions for access requests.
10.......................  Deadline for submitting requests for access
                            to Sensitive Unclassified Non-Safeguards
                            Information (SUNSI) with information:
                            supporting the standing of a potential party
                            identified by name and address; describing
                            the need for the information in order for
                            the potential party to participate
                            meaningfully in an adjudicatory proceeding.
60.......................  Deadline for submitting petition for
                            intervention containing: (i) Demonstration
                            of standing; (ii) all contentions whose
                            formulation does not require access to SUNSI
                            (+25 Answers to petition for intervention;
                            +7 requestor/petitioner reply).
20.......................  Nuclear Regulatory Commission (NRC) staff
                            informs the requestor of the staff's
                            determination whether the request for access
                            provides a reasonable basis to believe
                            standing can be established and shows need
                            for SUNSI. (NRC staff also informs any party
                            to the proceeding whose interest independent
                            of the proceeding would be harmed by the
                            release of the information.) If NRC staff
                            makes the finding of need for SUNSI and
                            likelihood of standing, NRC staff begins
                            document processing (preparation of
                            redactions or review of redacted documents).
25.......................  If NRC staff finds no ``need'' or no
                            likelihood of standing, the deadline for
                            requestor/petitioner to file a motion
                            seeking a ruling to reverse the NRC staff's
                            denial of access; NRC staff files copy of
                            access determination with the presiding
                            officer (or Chief Administrative Judge or
                            other designated officer, as appropriate).
                            If NRC staff finds ``need'' for SUNSI, the
                            deadline for any party to the proceeding
                            whose interest independent of the proceeding
                            would be harmed by the release of the
                            information to file a motion seeking a
                            ruling to reverse the NRC staff's grant of
                            access.
30.......................  Deadline 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.
A........................  If 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 + 3....................  Deadline for filing executed Non-Disclosure
                            Affidavits. Access provided to SUNSI
                            consistent with decision issuing the
                            protective order.
A + 28...................  Deadline for submission of contentions whose
                            development depends upon access to SUNSI.
                            However, if more than 25 days remain between
                            the petitioner's receipt of (or access to)
                            the information and the deadline for filing
                            all other contentions (as established in the
                            notice of hearing or opportunity for
                            hearing), the petitioner may file its SUNSI
                            contentions by that later deadline.
A + 53...................  (Contention receipt +25) Answers to
                            contentions whose development depends upon
                            access to SUNSI.
A + 60...................  (Answer receipt +7) Petitioner/Intervenor
                            reply to answers.
>A + 60..................  Decision on contention admission.
------------------------------------------------------------------------


[[Page 33252]]

[FR Doc. 2012-13426 Filed 6-4-12; 8:45 am]
BILLING CODE 7590-01-P