Notice; Applications and Amendments to Facility Operating Licenses Involving Proposed No Significant Hazards Considerations and Containing Sensitive Unclassified Non-Safeguards Information or Safeguards Information and Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information or Safeguards Information, 4247-4257 [E9-1152]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 14 (Friday, January 23, 2009)]
[Notices]
[Pages 4247-4257]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-1152]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0004]


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

I. Background

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

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

    The Commission has made a proposed determination that the following 
amendment requests involve

[[Page 4248]]

no significant hazards consideration. Under the Commission's 
regulations in 10 CFR 50.92, this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, TWB-05-B01M, Division of Administrative 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Documents may be examined, 
and/or copied for a fee, at the NRC(s Public Document Room (PDR), 
located at One White Flint North, Public File Area O1 F21, 11555 
Rockville Pike (first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, or 
at https://www.nrc.gov/reading-rm/doc-collections/cfr/part002/part002-
0309.html. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm.html. If a request for a hearing or petition for 
leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve all adjudicatory documents 
over the internet, or in some cases to mail copies on electronic 
storage media. Participants may not submit paper copies of their 
filings unless they seek

[[Page 4249]]

a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearing.docket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms 
Viewer(tm) to access the Electronic Information Exchange 
(EIE), a component of the E-Filing system. The Workplace Forms 
Viewer(tm) is free and is available at https://www.nrc.gov/
site-help/e-submittals/install-viewer.html. Information about applying 
for a digital ID certificate is available on NRC's public Web site at 
https://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC electronic filing 
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, 
Monday through Friday. The electronic filing Help Desk can be contacted 
by telephone at 1-866-672-7640 or by e-mail at MSHD.Resource@nrc.gov.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) FIRst class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition and/
or request should be granted and/or the contentions should be admitted, 
based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-
(viii).
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/ehd_proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr.resource@nrc.gov.

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

    Date of amendment request: November 20, 2008.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment revises Technical Specification (TS) 5.6.5, ``Core Operating 
Limits Report (COLR),'' to add a reference to an analytical method that 
will be used to determine core operating limits. The new reference, 
NEDC-33383P, ``GEXL97 Correlation Applicable to ATRIUM-10 Fuel,'' will 
allow the licensee to use a Global Nuclear Fuel method to determine 
fuel assembly critical power of AREVA ATRIUM-10 fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Core operating limits are established each operating cycle in 
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core 
Operating Limits Report (COLR).'' These core operating limits ensure 
that the fuel design limits are not exceeded during any conditions 
of normal operation or in the event of any Anticipated Operational 
Occurrence (AOO). The methods used to determine the operating limits 
are those previously found acceptable by the NRC and listed in TS 
section 5.6.5.b.
    A change to TS 5.6.5.b is requested to include an additional 
reference to the list of analytical methods. RBS [River Bend 
Station] currently operates with a full core of AREVA ATRIUM-10 fuel 
but is scheduled to load GE14 fuel during the next refueling outage.

[[Page 4250]]

RBS plans to use the analysis methods of the new fuel vendor, GNF 
[Global Nuclear Fuel], for the analysis of the mixed core. The 
GEXL97 correlation accurately models predicted core behavior and 
appropriately determines the overall critical power uncertainty of 
this method. In addition, the GEXL97 application range covers the 
range of expected operation of the ATRIUM-10 fuel during normal 
steady state and transient conditions in the RBS reload cores.
    The requested TS changes concern the use of analytical methods 
and do not involve any plant modifications or operational changes 
that could affect any postulated accident precursors or accident 
mitigation systems and do not introduce any new accident initiation 
mechanisms. The proposed changes have no effect on the type or 
amount of radiation released and [have] no effect on predicted 
offsite doses in the event of an accident. Thus, the proposed change 
does not affect the probability of an accident previously evaluated 
nor does it increase the radiological consequences of any accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed TS changes will not change the design function, 
reliability, performance, or operation of any plant systems, 
components, or structures. It does not create the possibility of a 
new failure mechanism, malfunction, or accident initiators not 
considered in the design and licensing bases. Plant operation will 
continue to be within the core operating limits that are established 
using NRC approved methods that are applicable to the RBS design and 
the RBS fuel.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds GEXL97 to the list of analytical 
methods in TS 5.6.5.b that can be used to determine core operating 
limits. Use of the GEXL97 correlation analytical method provides an 
equivalent level of protection as that currently provided. The 
change does not alter any method of analysis as described in the NRC 
approved versions of GESTAR-II [NEDE-24011-P-A, ``General Electric 
Standard Application for Reactor Fuel (GESTAR-II)'']. The proposed 
change does not modify the safety limits or setpoints at which 
protective actions are initiated, and do not change the requirements 
governing operation or availability of safety equipment assumed to 
operate to preserve the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: December 16, 2008.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). This 
amendment request would revise the Technical Specifications (TSs) 
Section 2.1.2, Safety Limit Minimum Critical Power Ratio (SLMCPR) for 
two-loop and single-loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes to Technical Specification do not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed Safety Limit MCPR (SLMCPR), and its use to 
determine the Operating Cycle 18 thermal limits, have been derived 
using NRC approved methods specified in the Reference section of the 
Technical Specification Bases Section for 2.0 SAFETY LIMITS. These 
methods do not change the method of operating the plant and have no 
effect on the probability of an accident initiating event or 
transient.
    The basis of the SLMCPR is to ensure no mechanistic fuel damage 
is calculated to occur if the limit is not violated. The new SLMCPR 
preserves the margin to transition boiling, and the probability of 
fuel damage is not increased.
    Therefore, the proposed changes to Technical Specifications do 
not involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes to Technical Specifications do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed changes result only from revised methods of 
analysis for the Cycle 18 core reload. These methods have been 
reviewed and approved by the NRC, do not involve any new or 
unapproved method for operating the facility, and do not involve any 
facility modifications. No new initiating events or transients 
result from these changes.
    Therefore, the proposed changes to technical specifications do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed changes to Technical Specifications do not 
involve a significant reduction in a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. The new SLMCPR was derived using NRC approved methods which 
are in accordance with the current fuel design and licensing 
criteria. The SLMCPR remains high enough to ensure that greater than 
99.9% of all fuel rods in the core will avoid transition boiling if 
the limit is not violated, thereby preserving the fuel cladding 
integrity.
    Therefore, the proposed changes to technical specifications do 
not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: July 25, 2008.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendments would revise Technical Specification 3.3.1.1, ``Reactor 
Protection System (RPS) Instrumentation,'' Surveillance Requirement 
(SR) 3.3.1.1.8 and TS 3.3.1.3, ``Oscillation Power Range Monitor (OPRM) 
Instrumentation,'' SR 3.3.1.3.2 to increase the frequency interval 
between local power range monitor calibrations from 1000 effective full 
power hours (EFPH) to 2000 EFPH for the LaSalle County Station, Units 1 
and 2 (LSCS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 4251]]

    Response: No.
    The proposed change is a result of increasing the surveillance 
interval of the LPRM [Local Power Range Monitor] calibration 
frequency from 1000 EFPH to 2000 EFPH. Increasing the frequency 
interval between required LRPM calibrations is acceptable due to 
improvements in the fuel analytical bases and therefore, the revised 
surveillance interval continues to ensure that the LPRM detector 
signal is adequately calibrated. Extending the LPRM calibration 
surveillance interval will increase the LPRM signal uncertainty 
value used in the LSCS SLMCPR [Safety Limit Minimum Critical Power 
Ratio] analysis, however, this increase in the LRPM signal 
uncertainty value is acceptable since the increase is bounded by the 
values used by the AREVA analysis.
    This change will not alter the operation of process variables, 
structures, systems, or components as described in the LSCS Updated 
Final Safety Analysis Report (UFSAR). The proposed change does not 
alter the initiation conditions or operational parameters for the 
system and there is no new equipment introduced by the extension of 
the LPRM calibration frequency interval. The performance of the 
Average Power Range Monitor (APRM), Rod Block Monitor (RBM) and 
Oscillation Power Range Monitor (OPRM) systems are not significantly 
affected by the proposed surveillance interval increase. The 
proposed LPRM calibration interval extension will have no 
significant effect on the Reactor Protection System (RPS) 
instrumentation accuracy during power maneuvers or transients and 
will, therefore, not significantly affect the performance of the 
RPS. As such, the probability of occurrences for a previously 
evaluated accident is not increased.
    The radiological consequences of an accident can be affected by 
the thermal limits existing at the time of the postulated accident, 
however, increasing the surveillance interval frequency will not 
increase the calculated thermal limits since all uncertainties 
associated with the increased interval are currently implemented and 
are currently used to calculate the existing Safety Limits. Plant 
specific evaluation of LPRM sensitivity to exposure has determined 
that the extended calibration frequency increases the LPRM signal 
uncertainty value used in the LSCS SLMCPR analysis, however, the 
increase is bounded by the values currently used in the safety 
analysis. Therefore, the thermal limit calculation is not 
significantly affected by LPRM calibration frequency, and thus the 
radiological consequences of any accident previously evaluated are 
not increased.
    Based on the above information, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The performance of the APRM, RBM, and OPRM systems are not 
significantly affected by the proposed LPRM surveillance interval 
increase. The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. For the proposed LPRM extended calibration 
interval frequency all uncertainties remain less than the 
uncertainties assumed in the existing thermal limit calculations. 
The proposed change does not change or introduce any new equipment, 
modes of system operation or failure mechanisms.
    Based on the above information, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change has no impact on equipment design or 
fundamental operation, and there are no changes being made to safety 
limits or safety system allowable values that would adversely affect 
plant safety as a result of the proposed LPRM surveillance interval 
increase. The performance of the APRM, RBM, and OPRM systems are not 
significantly affected by the proposed change. The margin of safety 
can be affected by the thermal limits existing at the time of the 
postulated accident; however, uncertainties associated with LPRM 
chamber exposure have no significant effect on the calculated 
thermal limits. Plant specific evaluation of LPRM sensitivity to 
exposure has determined that the extended calibration frequency 
increases the LPRM signal uncertainty value used in the LSCS SLMCPR 
analysis, however, the increase is bounded by the values currently 
used in the safety analysis. The thermal limit calculation is not 
significantly affected since the LPRM sensitivity with exposure is 
well defined. LPRM accuracy remains within the total nodal power 
uncertainty assumed in the thermal analysis basis, therefore 
maintaining thermal limits and the safety margin. The proposed 
change does not affect safety analysis assumptions or initial 
conditions and the margin of safety in the original safety analysis 
are therefore maintained.
    Based on this information, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

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

    Date of amendment request: October 13, 2008.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would revise the licensing basis by approving adoption of the 
Alternative Source Term (AST), in accordance with 10 CFR 50.67, for use 
in calculating the loss-of-coolant accident (LOCA) dose consequences. 
The proposed amendment would revise the Technical Specifications (TSs) 
to (1) change the TS definition for DOSE EQUIVALENT I-131 to adopt 
Federal Guidance Report (FGR) 11 dose conversion factors, (2) require 
operability of the Standby Liquid Control (SLC) system in Mode 3, to 
reflect its credit in the LOCA analysis, (3) establish a Main Steam 
(MS) Pathway leakage limit that effectively increases the previous MS 
isolation valve leakage limit, and (4) change TS Section 5.5.12 to 
reflect a requested permanent exemption from the requirements of 10 CFR 
Part 50, Appendix J, Option B, Paragraph III.A, to allow exclusion of 
MS Pathway leakage from the overall integrated leakage rate measured 
during the performance of a Type A test, and from the requirements of 
Appendix J, Option B, Paragraph III.B, to allow exclusion of the MS 
Pathway leakage from the combined leakage rate of the penetrations and 
valves subject to Type B and C tests.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    For the postulated design basis accident (DBA) LOCA, the AST is 
an input to the calculations that evaluate the radiological 
consequences of a LOCA. The AST and the requested Appendix J 
exemption do not affect the design of the plant or the manner in 
which the plant is normally operated. Adoption of the AST and the 
requested Appendix J exemption do not affect the initiators of a 
DBA. Neither the AST nor the requested Appendix J exemption [sic] 
affect the response to the DBA LOCA, or the pathway of the radiation 
released from the nuclear fuel. Rather, the AST better represents 
the physical characteristics of the radiation release.
    Because the initiators of a DBA are not affected by adoption of 
the AST for LOCA dose assessment, the probability of an accident are 
not increased by the proposed amendment or requested Appendix J 
exemption.
    The AST is an input to calculations used to evaluate the 
radiological consequences of

[[Page 4252]]

the LOCA. Use of the AST does not affect the plant response to the 
accident, or the pathways to the environment for the radiation and 
activity released from the fuel. The LOCA radiological analyses have 
been performed using the AST. Adoption of the AST methodology 
revises the acceptance criteria for the accident to the limits 
specified in 10 CFR 50.67. The results of those analyses demonstrate 
that the dose consequences are within the acceptance criteria 
presented in 10 CFR 50.67 and in NRC RG [Regulatory Guide] 1.183.
    Implementation of the AST for the LOCA involves the use of the 
SLC System to control the pH of the suppression pool during 
mitigation of a LOCA. As a result the proposed amendment revises the 
CNS [Cooper Nuclear Station] TS for the SLC System. These changes do 
not require any physical modification of the plant, nor result in 
any change in normal plant operation. This additional use of the SLC 
system does not compromise or adversely affect the function of the 
SLC system as a means of shutting down the reactor in addition to 
the control rods.
    Therefore, it is concluded that adoption of AST and granting of 
the Appendix J exemption do not involve a significant increase in 
the consequences of an accident previously evaluated. Based on the 
above discussion, it is concluded that the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Implementation of the LOCA AST and the requested Appendix J 
exemption do not involve a physical alteration of the plant or a 
change in how the plant is normally operated. No new or different 
types of equipment will be installed and there are no physical 
modifications to existing equipment associated with the proposed 
changes. The proposed changes, effectively increasing the allowable 
MSIV leakage, establishing a leakage limit for the MS Pathway, and 
crediting the SLC system for LOCA mitigation do not create 
initiators or precursors of a new or different kind of accident. New 
equipment or personnel failure modes that might initiate a new type 
of accident are not created as a result of the proposed amendment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment involves the implementation of a new 
licensing basis for the design basis LOCA. Approval of this change 
from the original source term to an AST, derived in accordance with 
the guidance of RG 1.183, results in revised acceptance criteria for 
the LOCA analysis. For the LOCA, RG 1.183 sets the Exclusion Area 
Boundary (EAB), Low Population Zone (LPZ), and Control Room limit 
consistent with 10 CFR 50.67. The AST LOCA radiological analysis has 
been performed using conservative methodologies, as specified in RG 
1.183. Safety margins have been evaluated and confirmed to have not 
been reduced. Analytical conservatism has been utilized to ensure 
that the analysis adequately bounds the limiting postulated event. 
The dose consequences of the DBA LOCA remain within the acceptance 
criteria presented in 10 CFR 50.67 and RG 1.183.
    The proposed changes continue to ensure that the doses at the 
EAB and LPZ boundary, as well as the Control Room, are within the 
corresponding regulatory limits.
    Since the proposed amendment continues to ensure the doses at 
the EAB, LPZ and Control Room are within corresponding regulatory 
limits, the proposed license amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: November 5, 2008.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The licensee 
proposed to increase the current maximum power level authorized by 
section 2.C(1) of the renewed facility operating license from 1,775 
megawatts thermal (Mwt) to 2,004 Mwt, an approximately 13 percent 
increase from the current licensed thermal power. The current maximum 
power level of 1,775 Mwt was approved in 1998, an increase of 6.3 
percent from the original licensed thermal power of 1670 Mwt. Thus, 
when approved, the licensee's proposed amendment would take the maximum 
power level to about 20 percent above the original license thermal 
power. The licensee's application addresses in details each of the 
following major technical areas: extended power uprate, containment 
analysis methods change, credit for containment overpressure for low 
head emergency core cooling system (ECCS) pumps, and reactor internal 
pressure differentials (RIPDs) for the steam dryer.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC). 
The licensee's NSHC analysis, addressing each technical area listed 
above, is reproduced below:

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

Extended Power Uprate

    Response: No.
    The probability (frequency of occurrence) of [d]esign [b]asis 
[a]ccidents occurring is not affected by the increased power level, 
because Monticello Nuclear Generating Plant (MNGP) continues to 
comply with the regulatory and design basis criteria established for 
plant equipment. A probabilistic risk assessment demonstrates that 
the calculated core damage frequencies do not significantly change 
due to [e]xtended [p]ower [u]prate (EPU). Scram setpoints (equipment 
settings that initiate automatic plant shutdowns) are established 
such that there is no significant increase in scram frequency due to 
EPU. No new challenges to safety-related equipment result from EPU.
    The changes in consequences of postulated accidents, which would 
occur from 102 percent of the EPU rated thermal power (RTP) compared 
to those previously evaluated, are acceptable. The results of EPU 
accident evaluations do not exceed the NRC[-]approved acceptance 
limits. The spectrum of postulated accidents and transients has been 
investigated, and are shown to meet the plant's currently licensed 
regulatory criteria. In the area of fuel and core design, for 
example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and 
other applicable Specified Acceptable Fuel Design Limits (SAFDL) are 
still met. Continued compliance with the SLMCPR and other SAFDLs 
will be confirmed on a cycle[-]specific basis consistent with the 
criteria accepted by the NRC.
    Challenges to the [r]eactor [c]oolant [p]ressure [b]oundary were 
evaluated at EPU conditions (pressure, temperature, flow, and 
radiation) and were found to meet their acceptance criteria for 
allowable stresses and overpressure margin.
    Challenges to the containment have been evaluated, and the 
containment and its associated cooling systems continue to meet the 
current licensing basis. The increase in the calculated post[-] LOCA 
suppression pool temperature above the currently assumed peak 
temperature was evaluated and determined to be acceptable. 
Radiological release events (accidents) have been evaluated, and 
have been shown to meet the guidelines of 10 CFR 50.67.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, variable RHR [residual heat 
removal] heat exchanger capability K-value, and mechanistic heat and

[[Page 4253]]

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

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

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

Reactor Internal Pressure Differentials (RIPDs) for the Steam Dryer

    Response: No.
    The revised steam dryer RIPDs are used in evaluating loads in 
reactor vessel internals for various conditions (i.e., during 
normal, upset and faulted conditions). The values more accurately 
represent the actual plant configuration. No plant structure, 
system, or component (SSC) is physically affected by the updated and 
expanded analyses. No method of operation of any plant SSC is 
affected. Therefore, there is no significant increase in the 
probability or consequence of a previously evaluated accident.
    The analyses supporting the above evaluations were performed at 
the EPU power level of 2,004 Mwt.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
     2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

Extended Power Uprate

    Response: No.
    Equipment that could be affected by EPU has been evaluated. No 
new operating mode, safety-related equipment lineup, accident 
scenario, or equipment failure mode was identified. The full 
spectrum of accident considerations has been evaluated and no new or 
different kind of accident has been identified. EPU uses developed 
technology and applies it within capabilities of existing or 
modified plant safety[-]related equipment in accordance with the 
regulatory criteria (including NRC[-]approved codes, standards and 
methods). No new accidents or event precursors have been identified.
    The MNGP TS require revision to implement EPU. The revisions 
have been assessed and it was determined that the proposed change 
will not introduce a different accident than that previously 
evaluated. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, variable RHR heat exchanger 
capability K-value, and mechanistic heat and transfer from the 
suppression pool surface to the wetwell airspace after 30 seconds 
for the long[-]term DBA-LOCA containment analysis are not relevant 
to accident initiation, but pertain to the method used to evaluate 
currently postulated accidents. The use of these analytical tools 
does not involve any physical changes to plant structures or 
systems, and does not create a new initiating event for the spectrum 
of events currently postulated. Further, they do not result in the 
need to postulate any new accident scenarios. Therefore, the 
containment analysis method change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

Credit for Containment Overpressure for Low Head ECCS Pumps

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

Reactor Internal Pressure Differentials for the Steam Dryer

    Response: No.
    The revised steam dryer RIPDs are used in evaluating loads in 
reactor vessel internals for various conditions (i.e., during 
normal, upset and faulted conditions). The steam dryer RIPDs are not 
relevant to accident initiation, but only pertain to the method used 
to evaluate reactor vessel internals loads. The revised steam dryer 
RIPD values more accurately represent the actual plant 
configuration. Therefore, the change to steam dryer RIPDs does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The analyses supporting the above evaluations were performed at 
the EPU power level of 2,004 Mwt.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
     3. Does the proposed change involve a significant reduction in 
a margin of safety?

Extended Power Uprate

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

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, variable RHR heat exchanger 
capability K-value, and mechanistic heat and mass transfer from the 
suppression pool surface to the wetwell airspace after 30 seconds 
for the long[-]term DBA-LOCA containment analysis are realistic 
phenomena and provide a conservative prediction of the plant 
response to DBA-LOCAs. The increase in pressure and temperature are 
relatively small and are within design limits. Therefore, the 
containment analysis methods change does not involve a significant 
reduction in the margin of safety.

Increase in Credit for Containment Overpressure for Low Head ECCS 
Pumps

    Response: No.
    The proposed changes revise containment response analytical 
methods and scope for containment pressure to assist in ECCS pump 
net positive suction head (NPSH). The changes are still based on 
conservative but more realistic analysis of available containment 
overpressure determined using analysis methods that minimize 
containment pressure and maximize suppression pool temperature. 
These changes do not constitute a significant reduction in the 
margin of safety.

Reactor Internal Pressure Differentials for the Steam Dryer

    Response: No.
    The revised steam dryer RIPDs are used in evaluating loads in 
reactor vessel internals for various conditions (i.e., during 
normal, upset and faulted conditions). The revised steam dryer RIPD 
values more accurately represent the actual plant configuration. The 
changes are still conservative but more accurately represent the 
MNGP configuration. These changes do not constitute a significant 
reduction in the margin of safety.
    The analyses supporting the above evaluations were performed at 
the EPU power level of 2,004 Mwt.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.


[[Page 4254]]


The NRC staff has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
proposed amendment involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Lois M. James.

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

    Date of amendment request: October 30, 2008.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). This 
amendment request would revise PPL Susquehanna, LLC, Unit 2 (PPL) 
Technical Specifications (TSs) Section 2.1.1.2, Minimum Critical Power 
Ratio Safety Limits (MCPRSLs) for two-loop and single-loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the two-loop and single-loop MCPRSLs do 
not directly or indirectly affect any plant system, equipment, 
component, or change the processes used to operate the plant. 
Further, the proposed MCPRSLs were generated using NRC approved 
methodology and meet the applicable acceptance criteria. Thus, this 
proposed amendment does not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    Prior to the startup of U2C15, licensing analyses are performed 
(using NRC approved methodology referenced in TS Section 5.6.5.b) to 
determine changes in the CPR as a result of anticipated operational 
occurrences. These results are added to the MCPRSL values to 
generate the MCPROLs in the COLR [Core Operating Limits Report]. 
These limits could be different from those specified for the 
previous Unit 2 COLR. The COLR operating limits thus assure that the 
MCPRSL will not be exceeded during normal operation or AOOs 
[anticipated operational occurrences]. Postulated accidents are also 
analyzed prior to the startup and the results shown to be within the 
NRC approved criteria.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to the two-loop and single-loop MCPRSLs do not 
directly or indirectly affect any plant system, equipment, or 
component and therefore does not affect the failure modes of any of 
these items. Thus, the proposed change does not create the 
possibility of a previously unevaluated operator error or a new 
single failure.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Since the proposed changes do not alter any plant system, 
equipment, component, or processes used to operate the plant, the 
proposed change will not jeopardize or degrade the function or 
operation of any plant system or component governed by TS. The 
proposed two-loop and single-loop MCPRSLs do not involve a 
significant reduction in the margin of safety as currently defined 
in the Bases of the applicable TS sections, because the proposed 
MCPRSLs preserve the required margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Mark G. Kowal.

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

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

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

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

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

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

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

     1. This order contains instructions regarding how potential 
parties to the proceedings listed above may request access to documents 
containing sensitive unclassified information (SUNSI and SGI).
     2. Within ten (10) days after publication of this notice of 
opportunity for hearing, any potential party as defined in 10 CFR 2.4 
who believes access to SUNSI or SGI is necessary for a response to the 
notice may request access to SUNSI or SGI. A ``potential party'' is any 
person who intends or may intend to participate as a party by 
demonstrating standing and the filing of an admissible contention under 
10 CFR 2.309. Requests submitted later than ten (10) days will not be 
considered absent a showing of good cause for the late filing, 
addressing why the request could not have been filed earlier.
    3. The requester shall submit a letter requesting permission to 
access SUNSI and/or SGI to the Office of the Secretary, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemakings and Adjudications Staff, and provide a copy to the 
Associate General Counsel for Hearings, Enforcement and Administration, 
Office of the General Counsel, Washington, DC 20555-0001. The expedited 
delivery or courier mail address for both offices is U.S. Nuclear 
Regulatory Commission, 11555 Rockville Pike, Rockville, MD 20852. The 
e-mail address for the Office of the Secretary and the Office of the

[[Page 4255]]

General Counsel are hearing.docket@nrc.gov and 
ogcmailcenter.resource@nrc.gov, respectively.\1\ The request must 
include the following information:
---------------------------------------------------------------------------

    \1\ See footnote 6. While a request for hearing or petition to 
intervene in this proceeding must comply with the filing 
requirements of the NRC's ``E-Filing Rule,'' the initial request to 
access SUNSI and/or SGI under these procedures should be submitted 
as described in this paragraph.
---------------------------------------------------------------------------

    a. A description of the licensing action with a citation to this 
Federal Register notice of opportunity for hearing;
    b. The name and address of the potential party and a description of 
the potential party's particularized interest that could be harmed by 
the action identified in (a);
    c. If the request is for SUNSI, the identity of the individual 
requesting access to SUNSI and the requester's need for the information 
in order to meaningfully participate in this adjudicatory proceeding, 
particularly why publicly available versions of the application would 
not be sufficient to provide the basis and specificity for a proffered 
contention;
     d. If the request is for SGI, the identity of the individual 
requesting access to SGI and the identity of any expert, consultant or 
assistant who will aid the requester in evaluating the SGI, and 
information that shows:
    (i) Why the information is indispensable to meaningful 
participation in this licensing proceeding; and
    (ii) The technical competence (demonstrable knowledge, skill, 
experience, training or education) of the requester to understand and 
use (or evaluate) the requested information to provide the basis and 
specificity for a proffered contention. The technical competence of a 
potential party or its counsel may be shown by reliance on a qualified 
expert, consultant or assistant who demonstrates technical competence 
as well as trustworthiness and reliability, and who agrees to sign a 
non-disclosure affidavit and be bound by the terms of a protective 
order; and
    e. If the request is for SGI, Form SF-85, ``Questionnaire for Non-
Sensitive Positions,'' Form FD-258 (fingerprint card), and a credit 
check release form completed by the individual who seeks access to SGI 
and each individual who will aid the requester in evaluating the SGI. 
For security reasons, Form SF-85 can only be submitted electronically, 
through a restricted-access database. To obtain online access to the 
form, the requester should contact the NRC's Office of Administration 
at 301-492-3524.\2\ The other completed forms must be signed in 
original ink, accompanied by a check or money order payable in the 
amount of $191.00 to the U.S. Nuclear Regulatory Commission for each 
individual, and mailed to the: Office of Administration, Security 
Processing Unit, Mail Stop TWB-05-B32M, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0012.
---------------------------------------------------------------------------

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

    These forms will be used to initiate the background check, which 
includes fingerprinting as part of a criminal history records check. 
Note: copies of these forms do not need to be included with the request 
letter to the Office of the Secretary, but the request letter should 
state that the forms and fees have been submitted as described above.
    4. To avoid delays in processing requests for access to SGI, all 
forms should be reviewed for completeness and accuracy (including 
legibility) before submitting them to the NRC. Incomplete packages will 
be returned to the sender and will not be processed.
    5. Based on an evaluation of the information submitted under items 
2 and 3.a through 3.d, above, the NRC staff will determine within ten 
days of receipt of the written access request whether (1) there is a 
reasonable basis to believe the petitioner is likely to establish 
standing to participate in this NRC proceeding, and (2) there is a 
legitimate need for access to SUNSI or need to know the SGI requested. 
For SGI, the need to know determination is made based on whether the 
information requested is necessary (i.e., indispensable) for the 
proposed recipient to proffer and litigate a specific contention in 
this NRC proceeding \3\ and whether the proposed recipient has the 
technical competence (demonstrable knowledge, skill, training, 
education, or experience) to evaluate and use the specific SGI 
requested in this proceeding.
---------------------------------------------------------------------------

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

    6. If standing and need to know SGI are shown, the NRC staff will 
further determine based upon completion of the background check whether 
the proposed recipient is trustworthy and reliable. The NRC staff will 
conduct (as necessary) an inspection to confirm that the recipient's 
information protection systems are sufficient to protect SGI from 
inadvertent release or disclosure. Recipients may opt to view SGI at 
the NRC's facility rather than establish their own SGI protection 
program to meet SGI protection requirements.
    7. A request for access to SUNSI or SGI will be granted if:
    a. The request has demonstrated that there is a reasonable basis to 
believe that a potential party is likely to establish standing to 
intervene or to otherwise participate as a party in this proceeding;
    b. The proposed recipient of the information has demonstrated a 
need for SUNSI or a need to know for SGI, and that the proposed 
recipient of SGI is trustworthy and reliable;
    c. The proposed recipient of the information has executed a Non-
Disclosure Agreement or Affidavit and agrees to be bound by the terms 
of a Protective Order setting forth terms and conditions to prevent the 
unauthorized or inadvertent disclosure of SUNSI and/or SGI; and
    d. The presiding officer has issued a protective order concerning 
the information or documents requested.\4\ Any protective order issued 
shall provide that the petitioner must file SUNSI or SGI contentions 25 
days after receipt of (or access to) that information. However, if more 
than 25 days remain between the petitioner's receipt