Notice; Applications and Amendments to Facility Operating 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, 458-467 [E9-31060]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 2 (Tuesday, January 5, 2010)]
[Notices]
[Pages 458-467]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-31060]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0564]


Notice; Applications and Amendments to Facility Operating 
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

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).

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 no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), Section 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 
and Directives Branch (RDB), TWB-05-

[[Page 459]]

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. Written comments may also be faxed to the RDB at 301-
492-3446. 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 requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, 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 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
hearing.docket@nrc.gov, or by telephone at (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 
(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 NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
https://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 EIE, users will be required to install a Web 
browser plug-

[[Page 460]]

in from the NRC 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 https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at https://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 e-mail notice confirming receipt of the document. The 
E-Filing 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 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 https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, or by a toll-free call at (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 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from January 5, 2010. 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).
    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 electronically 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.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant (HBRSEP) Unit No. 2, Darlington County, South Carolina

    Date of amendment request: June 19, 2009, as supplemented by letter 
dated October 20, 2009.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would revise TS 3.3.1, ``Reactor Protection System 
Instrumentation.'' The proposed change revises the requirements related 
to the reactor protection system interlock for the turbine trip input 
to the reactor protection system.
    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 Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed change provides revised requirements for the 
reactor protection system interlock associated with the turbine trip 
protection function. The proposed change will allow the interlock 
for turbine trip function to be raised from the current interlock 
setting of nominally 10 percent reactor power to nominally 40 
percent reactor power.
    This change will allow the reactor to continue operating safely 
at power levels up to nominally 40 percent when the turbine is not 
operating. The applicable accident analyses, as described in the 
HBRSEP, Unit No. 2, Updated Final Safety Analysis Report (UFSAR) 
have been reviewed. The turbine trip input to reactor trip has been 
verified to be either not used in the accident analyses or that the 
change does not adversely affect the analyses results and 
conclusion. Therefore, it is concluded that the consequences as 
described in the UFSAR accident analyses are unaffected by the 
proposed change.
    An analysis of plant response to a turbine trip at nominally 40 
percent power provided with the amendment request shows that the 
applicable acceptance criteria are met. Specifically, analysis has 
shown that a turbine trip without a reactor trip below 40 percent 
power does not challenge the pressurizer PORVs [power operated 
relief valves] or the steam generator safety valves; thereby, not 
adversely affecting the probability of a small break LOCA [loss of 
coolant accident] due to a stuck open PORV, or an excessive cooldown 
event due to a stuck open steam generator safety valve. As a result, 
the probability of any accident

[[Page 461]]

previously evaluated is not significantly increased by the proposed 
changes.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated.
    As described above, the proposed change provides revised 
requirements for the reactor protection system interlock associated 
with the turbine trip protection function. The proposed change will 
allow the interlock for turbine trip function to be raised from the 
current interlock setting of nominally 10 percent reactor power to 
nominally 40 percent reactor power.
    No new accident initiators or precursors are introduced by the 
proposed change. Changing the interlock for the reactor trip on 
turbine trip from P-7 to P-8 changes the power level associated with 
enabling and disabling the reactor trip on turbine trip function. 
The turbine pressure input to the reactor protection system 
permissive is not an accident initiator. The change does not affect 
how the associated trip functional units operate or function. The 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated because these interlock 
changes do not affect the way that the associated trip functional 
units operate or function.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety.
    As described above, the proposed change provides revised 
requirements for the reactor protection system interlock associated 
with the turbine trip protection function. The proposed change will 
allow the interlock for the turbine trip function to be raised from 
the current interlock setting of nominally 10 percent reactor power 
to nominally 40 percent reactor power.
    Also, as previously described, this change will allow the 
reactor to continue operating safely at power levels up to nominally 
40 percent when the turbine is not operating. The applicable UFSAR 
accident analyses have been reviewed and it is concluded that the 
accident analyses are unaffected by the proposed change. An analysis 
of plant response to a turbine trip at nominally 40 percent power 
shows that the applicable acceptance criteria are met. Based on 
these evaluations, the margins of safety that could potentially have 
been impacted by the proposed change associated with the reactor, 
which include departure from nucleate boiling (DNB) and fuel 
temperature margins, and the margin of safety associated with 
reactor coolant system integrity, are not affected.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: October 27, 2009.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). This 
amendment request would change the Technical Specifications to provide 
revised values for the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) for both single and dual recirculation 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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The basis of the Safety Limit MCPR (SLMCPR) is to ensure no 
mechanistic fuel damage is calculated to occur if the limit is not 
violated. The new SLMCPR values preserve the existing margin to 
transition boiling and probability of fuel damage is not increased. 
The derivation of the revised SLMCPR for Vermont Yankee for 
incorporation into the Technical Specifications and its use to 
determine plant and cycle-specific thermal limits has been performed 
using NRC approved methods. These plant-specific calculations are 
performed each operating cycle and if necessary, will require future 
changes to these values based upon revised core designs. The revised 
SLMCPR values do not change the method of operating the plant and 
have no effect on the probability of an accident initiating event or 
transient.
    Based on the above, Vermont Yankee has concluded that the 
proposed change will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes result only from a specific analysis for 
the Vermont Yankee core reload design. These changes do not involve 
any new or different methods for operating the facility. No new 
initiating events or transients result from these changes.
    Based on the above, Vermont Yankee has concluded that the 
proposed change will not create the possibility of a new or 
different kind of accident from those previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The new SLMCPR is calculated using NRC approved methods with 
plant and cycle specific parameters for the current core design. The 
SLMCPR value remains conservative 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. The operating MCPR limit is set appropriately above the 
safety limit value to ensure adequate margin when the cycle specific 
transients are evaluated. Accordingly, the margin of safety is 
maintained with the revised values.
    As a result, Vermont Yankee has determined that the proposed 
change will not result in 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: Nancy L. Salgado.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: October 27, 2009.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment revises the Technical Specifications to increase the two 
recirculation loop minimum critical power ratio (MCPR) safety limit 
from 1.08 to 1.09 and the single recirculation loop MCPR safety limit 
from 1.10 to 1.12.

[[Page 462]]

    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 Minimum Critical Power Ratio (MCPR) limit is defined in the 
Bases to Technical Specification 2.1.1.2 as that limit, ``that, in 
the event of an AOO [(Anticipated Operational Occurrence)] from the 
limiting condition of operation, at least 99.9% of the fuel rods in 
the core would be expected to avoid boiling transition.'' The MCPR 
safety limit satisfies the requirements of General Design Criterion 
10 of Appendix A to 10CFR50 regarding acceptable fuel design limits. 
The MCPR safety limit is reevaluated for each reload using NRC-
approved methodologies. The analyses for GGNS [Grand Gulf Nuclear 
Station] Cycle 18 have concluded that a two-loop MCPR safety limit 
of 1.09, based on the application of Global Nuclear Fuels' NRC 
approved MCPR safety limit methodology, will ensure that this 
acceptance criterion is met. For single-loop operation, a MCPR 
safety limit of 1.12, also ensures that this acceptance criterion is 
met. The MCPR operating limits are presented and controlled in 
accordance with the GGNS Core Operating Limits Report (COLR).
    The requested Technical Specification changes do not involve any 
plant modifications or operational changes that could affect system 
reliability or performance or that could affect the probability of 
operator error. The requested changes do not affect any postulated 
accident precursors, do not affect any accident mitigating systems, 
and do not introduce any new accident initiation mechanisms.
    Therefore, the changes to the Minimum Critical Power Ratio 
safety limit do not involve a significant increase in the 
probability or consequences of any 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 GNF2 fuel to be used in Cycle 18 is of a design compatible 
with the co-resident GE14 and ATRIUM-10 fuel. Therefore, the 
introduction of GNF2 fuel into the Cycle 18 core will not create the 
possibility of a new or different kind of accident. The proposed 
changes do not involve any new modes of operation, any changes to 
setpoints, or any plant modifications. The proposed revised MCPR 
safety limits have accounted for the mixed fuel core and have been 
shown to be acceptable for Cycle 18 operation. Compliance with the 
criterion for incipient boiling transition continues to be ensured. 
The core operating limits will continue to be developed using NRC 
approved methods which also account for the mixed fuel core design. 
The proposed MCPR safety limits or methods for establishing the core 
operating limits do not result in the creation of any new precursors 
to an accident.
    Therefore, this 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 MCPR safety limits have been evaluated in accordance with 
Global Nuclear Fuels NRC-approved cycle-specific safety limit 
methodology to ensure that during normal operation and during AOO's 
at least 99.9% of the fuel rods in the core are not expected to 
experience transition boiling. The proposed revised MCPR safety 
limits have accounted for the mixed fuel core and have been shown to 
be acceptable for Cycle 18 operation. Compliance with the criterion 
for incipient boiling transition continues to be ensured. On this 
basis, the implementation of the change to the MCPR safety limits 
does not involve a significant reduction in a 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 Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: November 3, 2009.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment revises the Technical Specifications (TSs) to reflect the 
installation of the digital General Electric--Hitachi Nuclear 
Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring 
(PRNM) System.
    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 (frequency of occurrence) of design basis 
accidents (DBAs) occurring is not affected by the NUMAC PRNM System, 
since the system does not interact with equipment whose failure 
could cause an accident. Compliance with the regulatory criteria 
established for plant equipment are maintained with the installation 
of the upgraded NUMAC PRNM System. Scram setpoints in the NUMAC PRNM 
System are established such that the analytical limits are met.
    The unavailability of the new NUMAC PRNM System is equal to or 
less than the existing system and, as a result, the scram 
reliability is equal to or better than the existing analog power 
system. No new challenges to safety-related equipment result from 
the NUMAC PRNM System modification. Therefore, the proposed change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The proposed change replaces the current Option E-I-A stability 
solution with an NRC-approved Option III long-term stability 
solution. The NUMAC PRNM hardware incorporates the Oscillation Power 
Range Monitor (OPRM) Option III detect-and-suppress solution, which 
has been previously reviewed and approved by the NRC. The OPRM meets 
[10 CFR Part 50, Appendix A] General Design Criterion (GDC) 10, 
Reactor Design, and GDC 12, Suppression of Reactor Power 
Oscillations, requirements by automatically detecting and 
suppressing design basis thermal-hydraulic oscillations prior to 
exceeding the fuel Minimum Critical Power Ratio (MCPR) Safety Limit.
    Based on the above, installation of the new NUMAC PRNM System 
with the OPRM Option III stability solution integrated into the 
NUMAC PRNM equipment does not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The components of the NUMAC PRNM System are equivalent or of 
better design and qualification criteria than those currently 
installed and utilized in the plant. No new operating mode, safety-
related equipment lineup, accident scenario, or interaction mode not 
reviewed and approved as part of the design and licensing of the 
NUMAC PRNM System has been identified. Therefore, the NUMAC PRNM 
System retrofit does not adversely affect plant equipment.
    The new NUMAC PRNM System uses digital equipment that has 
software-controlled digital processing compared to the existing 
power range system that uses mostly analog and discrete component 
processing. Specific failures of hardware and potential software 
common-cause failures are different from the existing system. The 
effects of potential software common-cause failure are mitigated by 
specific hardware design and

[[Page 463]]

system architecture as discussed in Section 6.0 of [GE Nuclear 
Energy Licensing Topical Report] NEDC-32410P-A. Failure(s) of the 
system have the same overall effect as the present design. No new or 
different kinds of accidents are introduced. Therefore, the NUMAC 
PRNM System does not adversely effect plant equipment.
    The currently installed Average Power Range Monitoring (APRM) 
system is replaced with a NUMAC PRNM System that performs the 
existing power range monitoring functions and adds an OPRM to react 
automatically to potential reactor thermal-hydraulic instabilities.
    Based on the above, 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 proposed TS changes associated with the NUMAC PRNM System 
retrofit implement the constraints of the NUMAC PRNM System design 
and related stability analyses. The NUMAC PRNM System change does 
not impact reactor operating parameters or the functional 
requirements of the APRM system. The replacement equipment continues 
to provide information, enforce control rod blocks, and initiate 
reactor scrams under appropriate specified conditions. The proposed 
change does not reduce safety margins. The replacement APRM 
equipment has improved channel trip accuracy compared to the current 
analog system, and meets or exceeds system requirements previously 
assumed in setpoint analysis. Thus, the ability of the new equipment 
to enforce compliance with margins of safety equals or exceeds the 
ability of the equipment which it replaces.
    Therefore, the proposed changes do not involve a 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.

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: October 5, 2009.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The 
amendment(s) would revise Technical Specification (TS) 4.3.1, 
``Criticality,'' to address a non-conservative TS. The proposed change 
addresses the Boraflex degradation issue in the LaSalle County Station 
(LSCS) Unit 2 spent fuel storage racks by revising TS Section 4.3.1 to 
allow the use of NETCO-SNAP-IN[reg] rack inserts in LSCS Unit 2 spent 
fuel storage rack cells as a replacement for the neutron absorbing 
properties of the existing Boraflex panels.
    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 adds an additional requirement to TS Section 
4.3.1 to install a NETCO-SNAP-IN[reg] rack insert in spent fuel 
storage rack cells that cannot otherwise maintain the requirements 
of TS Section 4.3.1.1.a to ensure that the effective neutron 
multiplication factor, Keff, is less than or equal to 
0.95, if the spent fuel pool (SFP) is fully flooded with unborated 
water. The proposed change also includes a revision to TS Section 
4.3.1 to specify the bounding reactivity fuel design allowed for 
storage in the Unit 1 and Unit 2 SFPs. Since the proposed change 
pertains only to the SFP, only those accidents that are related to 
movement and storage of fuel assemblies in the SFP could be 
potentially affected by the proposed change.
    The current licensing basis for the LSCS Unit 2 SFP credits the 
neutron absorbing properties of the Boraflex neutron poison material 
in the spent fuel storage racks. The current licensing basis 
demonstrates: (1) Adequate margin to criticality for spent fuel 
storage rack cells that credit the neutron absorption capabilities 
of Boraflex, (2) adequate margin for fuel assemblies inadvertently 
placed into locations adjacent to the spent fuel storage racks, and 
(3) adequate margin for assemblies accidentally dropped onto the 
spent fuel storage racks. Therefore, the probability that a 
misplaced fuel assembly would result in an inadvertent criticality 
is unchanged since the process and procedural controls governing 
fuel movement in the SFP will not be changed. The dose consequences 
of the most limiting drop of a fuel assembly in the SFP is limited 
by the number of the fuel rods damaged and other engineered features 
unaffected by the proposed change, including the fuel design, fuel 
decay time, water level in the SFP, water temperature of the SFP, 
and the engineering features of the Reactor Building Ventilation 
System.
    The installation of NETCO-SNAP-IN[reg] rack inserts does not 
result in a significant increase in the probability of an accident 
previously analyzed. The revised criticality analysis takes no 
credit for the Boraflex material. The use of a rack insert provides 
an alternative neutron absorber to take the place of the degraded 
Boraflex material, without removal of the existing Boraflex. The 
probability that a fuel assembly would be dropped is unchanged by 
the installation of the NETCO-SNAP-IN[reg] 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[reg] rack inserts does not 
result in a significant increase in the consequence of an accident 
previously analyzed. A criticality analysis has been prepared to 
demonstrate adequate margin to criticality for spent fuel storage 
rack cells with rack inserts in the LSCS Unit 2 SFP, and adequate 
criticality margin for assemblies accidentally dropped onto the 
spent fuel storage racks.
    The installation of NETCO-SNAP-IN@ rack inserts does not affect 
the consequences of a dropped fuel assembly. The consequences of 
dropping a fuel assembly onto any other fuel assembly or other 
structure are unaffected by the change. The consequences of dropping 
a fuel assembly onto a spent fuel storage rack cell with a rack 
insert are bounded by the event of dropping an assembly onto another 
assembly, both for criticality and for radiological consequences. 
For criticality, the effects on Keff of dropping a fuel 
assembly have been evaluated and are acceptable. For radiological 
consequences, the number of rods damaged when a fuel assembly is 
accidentally dropped onto a spent fuel storage rack cell with or 
without a rack insert is bounded by the number of rods damaged by an 
assembly dropped onto another assembly. The change does not affect 
the effectiveness of the other engineered design features to limit 
the offsite dose consequences of the limiting fuel assembly drop 
accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Onsite storage of spent fuel assemblies in the SFP is a normal 
activity for which LSCS 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 SFP, 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 storage rack 
cells that have been determined unusable based on the degradation of 
Boraflex within those spent fuel 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 Boraflex in that 
cell without the potential for degradation. These devices do not add 
any limiting structural loads or affect the removal of decay heat 
from

[[Page 464]]

the assemblies. No change in total heat load in the SFP is being 
made. The devices are resistant to corrosion and will maintain their 
structural integrity over the life of the SFP. An accidental fuel 
assembly drop does not challenge their structural integrity. 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 change does not create the possibility of a 
misloaded assembly into a spent fuel storage rack cell.
    The misloading of a more reactive assembly targeted for 
placement in the LSCS Unit 1 SFP or the LSCS Unit 2 SFP Boraflex 
region in a rack insert region of the LSCS Unit 2 SFP has been 
prevented since the most reactive fuel assembly at LSCS is bounded 
by the rack insert criticality analysis, and the most reactive fuel 
assembly allowed for future insertion in either the Unit 1 or Unit 2 
SFP is being limited to the reference bounding ATRIUM-10 fuel 
assembly.
    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.
    LSCS 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 an alternative 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 
analysis demonstrates 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 
margin of safety for 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 consequences of a 
dropped fuel assembly without a rack insert. The proposed change 
also maintains the capacity of the Unit 2 SFP to be no more than 
4078 fuel assemblies.
    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. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station, 
Unit No. 1, Salem County, New Jersey

    Date of amendment request: October 8, 2009.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would revise Technical Specification (TS) 6.8.4.i, ``Steam 
Generator (SG) Program,'' by adding a one-time alternate repair 
criterion that excludes certain portions of the tube below the top of 
the SG tubesheet from periodic SG tube inspections. In addition, the 
proposed amendment would revise TS 6.9.10, ``Steam Generator Tube 
Inspection Report,'' to provide reporting requirements specific to the 
alternate repair criteria. The proposed amendment is supported by 
Westinghouse Electric Company, LLC Topical Report WCAP-17071-P, ``H*: 
Alternate Repair Criteria for the Tubesheet Expansion Region in Steam 
Generators with Hydraulically Expanded Tubes (Model F).'' H* 
(pronounced ``H star'') is the length of hydraulically expanded SG tube 
that must remain intact within the tubesheet in order for the joint to 
resist pullout and leakage due to normal operating and accident 
conditions.
    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 do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    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.
    During 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. Based on this design, 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 6.8.4.i, 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 Technical 
Specification 6.8.4.i. 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 probability of a SLB is unaffected by the potential failure 
of a steam generator tube as the failure of tube is not an initiator 
for a SLB event.
    The leakage factor of 2.16 for Salem Unit 1, for a postulated 
SLB/FLB, has been calculated as shown in Table 9-7 of WCAP-17071-P 
as revised by the response to RAI [request for additional 
information] 24 (Attachment 7 [to the application dated October 8, 
2009]). Through application of the limited tubesheet inspection 
scope, the existing operating leakage limit provides assurance that 
excessive leakage (i.e., greater than accident analysis assumptions) 
will not occur. The accident analysis calculations have an 
assumption of 0.6 [gallons per minute (gpm)] at room temperature 
(gpmRT) primary-to-secondary leakage in a single SG and 1 gpm at 
room temperature (gpmRT) total primary-to-secondary leakage for all 
SGs. This apportioned primary-to-secondary leakage is used in the 
Main Steam Line Break and Locked Rotor accidents. Primary-to-
secondary leakage of 1 gpm at room temperature (gpmRT) in a single 
SG is used in the Control Rod Ejection (CRE) accident.
    No leakage factor will be applied to the locked rotor or control 
rod ejection transients due to their short duration.

[[Page 465]]

    The TS operational leak rate limit is 150 gallons per day (gpd) 
(0.104 gpmRT). The maximum accident leak rate ratio for Salem Unit 1 
is 2.16. Consequently, this results in significant margin between 
the conservatively estimated accident leakage and the allowable 
accident leakage.
    For the condition monitoring (CM) assessment, the component of 
leakage from the prior cycle from below the H* distance will be 
multiplied by a factor of 2.16 and added to the total leakage from 
any other source and compared to the allowable accident induced 
leakage limit. For the operational assessment (OA), 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 2.16 and compared to the observed 
operational leakage.
    Based on the above, 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.
    The proposed change that alters the steam generator 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.
    The proposed change defines the safety significant portion of 
the tube that must be inspected and repaired (plugged). WCAP-17071-P 
identifies the specific inspection depth below which any type tube 
degradation shown to have no impact on the performance criteria in 
[Nuclear Energy Institute (NEI) document] NEI 97-06 [Revision] 2, 
``Steam Generator Program Guidelines.''
    The proposed change that alters the steam generator inspection 
and reporting criteria maintains the required structural margins of 
the SG tubes for both normal and accident conditions. Nuclear Energy 
Institute 97-06, ``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-1707[1]-P shows that significant margin exists between an 
acceptable level of leakage during normal operating conditions that 
ensures meeting the accident-induced leakage assumptions and the TS 
leakage limit of 150 gpd.

    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: Vincent Zabielski, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant (HBRSEP) Unit No. 2, Darlington County, South Carolina

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

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

PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station, 
Unit No. 1, Salem County, New Jersey

    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.

[[Page 466]]

    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 e-mail 
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);
    (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 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 
either after a determination on standing and need for access, or after 
a determination on trustworthiness and reliability, 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 Gran