[Federal Register Volume 76, Number 21 (Tuesday, February 1, 2011)]
[Notices]
[Pages 5614-5626]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-2027]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2011-0021]


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, NRC, 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) 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.

[[Page 5615]]

    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, Rules, 
Announcements and Directives Branch (RADB), 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. 
Written comments may also be faxed to the RADB 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, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852-2738.
    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, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852-2738, or 
at http://www.nrc.gov/reading-rm/doc-collections/cfr/part002/part002-0309.html. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm.html. If a request for a hearing or petition for 
leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the 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, then any hearing held would take place before 
the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
<a href="/cdn-cgi/l/email-protection#19717c786b70777e377d767a727c6d59776b7a377e766f"><span class="__cf_email__" data-cfemail="fd95989c8f94939ad399929e969889bd938f9ed39a928b">[email protected]</span></a>, or by telephone at 301-415-1677, to request (1) 
a digital identification (ID) certificate, which allows the participant 
(or its counsel or representative) to digitally sign documents and 
access the E-Submittal server for any proceeding in which it is 
participating; and (2) advise the Secretary that the participant will 
be submitting a request or petition for hearing (even in instances in 
which the participant, or its counsel or representative, already holds 
an NRC-

[[Page 5616]]

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 http://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 
http://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an 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 http://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
<a href="/cdn-cgi/l/email-protection#a1ecf2e9e58ff3c4d2ced4d3c2c4e1cfd3c28fc6ced7"><span class="__cf_email__" data-cfemail="86cbd5cec2a8d4e3f5e9f3f4e5e3c6e8f4e5a8e1e9f0">[email protected]</span></a>, or by a toll-free call at 1-866- 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852-2738, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/EHD/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. With 
respect to copyrighted works, except for limited excerpts that serve 
the purpose of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    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, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852-2738. 
Publicly available records will be accessible electronically from the 
ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/adams.html. If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by e-mail to <a href="/cdn-cgi/l/email-protection#90e0f4e2bee2f5e3ffe5e2f3f5d0fee2f3bef7ffe6"><span class="__cf_email__" data-cfemail="88f8ecfaa6faedfbe7fdfaebedc8e6faeba6efe7fe">[email protected]</span></a>.

Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Units 2 and 3, New London County, 
Connecticut

    Date of amendment request: July 12, 2010, as supplemented by letter 
dated August 5, 2010.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The licensee 
proposed an amendment to the Facility Operating Licenses for Millstone 
Power Station, Units 2 and 3 (MPS2 and MPS3, respectively). This 
amendment request pertains to the MPS2 and MPS3 Cyber Security Plans. 
In the same amendment request letter, sent under Dominion Resources 
Services, Inc. (DRC) letterhead, Kewaunee Power Station, Surry Power 
Station Units 1 and 2, and North Anna Power Station Units 1 and 2, 
submitted amendment requests pertaining to their Cyber Security Plans. 
This notice only addresses the application as it pertains to MPS2 and 
MPS3. The licensee requested NRC approval of the MPS2 and MPS3 Cyber 
Security Plan, provided a proposed implementation schedule, and 
proposed to add a sentence to License Condition

[[Page 5617]]

2.C.4, ``Physical Protection,'' of MPS2, Facility Operating License 
(FOL) DPR-65 and to License Condition 2.E, of MPS3, FOL NPF-49, that 
would affirm when the licensee would fully implement and maintain in 
effect all provisions of the Cyber Security Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration (NSHC). The NRC 
staff reviewed the licensee's NSHC analysis against the standards of 10 
CFR 50.92(c). The NRC staff's review 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 Plan establishes the licensing basis for the Cyber Security 
Program for the sites. The Plan establishes how to achieve high 
assurance that specified nuclear power plant digital computer and 
communication systems, networks and functions are adequately 
protected against cyber attacks up to and including the design basis 
threat.
    Part one of the proposed change is designed to achieve high 
assurance that the systems are protected from cyber attacks. The 
Plan describes how plant modifications that involve digital computer 
systems are reviewed to provide high assurance of adequate 
protection against cyber attacks, up to and including the design 
basis threat. The proposed change does not alter accident analysis 
assumptions, add any initiators, or affect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected. The first part of the proposed 
change is designed to achieve high assurance that the systems within 
the scope of the requirement are protected from cyber attacks and 
has no impact on the probability or consequences of an accident 
previously evaluated. The proposed change implements a Cyber 
Security Plan as a requirement not formally addressed previously. As 
such, the proposed Plan provides a significant enhancement to cyber 
security where no requirement existed before.
    The second part of the proposed change adds a sentence to the 
existing facility license conditions for Physical Protection. These 
changes are administrative and have no impact on the probability or 
consequences of an accident previously evaluated.
    Therefore, it is concluded that these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This proposed amendment provides assurance that safety-related 
structures, systems and components (SSCs) are protected from cyber 
attacks. Implementation of 10 CFR 73.54 and the inclusion of a plan 
in the FOL do not result in the need of any new or different design-
basis accident analysis. It does not introduce new equipment that 
could create a new or different kind of accident, and no new 
equipment failure modes are created. As a result, no new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of this proposed amendment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is associated with the confidence in the 
ability of the fission product barriers (i.e., fuel cladding, 
reactor coolant pressure boundary, and containment structure) to 
limit the level of radiation to the public. The proposed amendment 
would not alter the way any safety-related SSC functions and would 
not alter the way the plant is operated. The amendment provides 
assurance that safety-related SSCs are protected from cyber attacks. 
The proposed amendment would not introduce any new uncertainties or 
change any existing uncertainties associated with any safety limit. 
The proposed amendment would have no impact on the structural 
integrity of the fuel cladding, reactor coolant pressure boundary, 
or containment structure. Based on the above considerations, the 
proposed amendment would not degrade the confidence in the ability 
of the fission product barriers to limit the level of radiation to 
the public.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc.,
    120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Harold K. Chernoff.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois Docket Nos. STN 
50-454 and 50-455, Byron Station, Units 1 and 2, Ogle County, Illinois

    Date of amendment request: December 14, 2010.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The 
amendment would revise Technical Specification (TS) 5.5.9, ``Steam 
Generator (SG) Program,'' to exclude portions of the tubes within the 
tubesheet from periodic SG inspections and plugging or repair. In 
addition, this amendment request proposes to revise TS 5.6.9, ``Steam 
Generator (SG) Tube Inspection Report,'' to remove reference to 
previous interim alternate repair criteria and provide reporting 
requirements specific to the temporary alternate criteria.
    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 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 various accidents previously evaluated, the proposed 
changes only affect the steam generator tube rupture (SGTR), 
postulated steam line break (SLB), feedwater line break (FLB), 
locked rotor and control rod ejection accident evaluations. Loss-of-
coolant accident (LOCA) conditions cause a compressive axial load to 
act on the tube. Therefore, since the LOCA tends to force the tube 
into the tubesheet rather than pull it out, it is not a factor in 
this amendment request. Another faulted load consideration is a safe 
shutdown earthquake (SSE); however, the seismic analysis of Model D5 
SGs has shown that axial loading of the tubes is negligible during 
an SSE.
    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 draft Regulatory Guide (RG) 1.121, 
``Bases for Plugging Degraded PWR Steam Generator Tubes,'' and TS 
5.5.9, 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

[[Page 5618]]

leakage integrity of the SG tubes consistent with the performance 
criteria of TS 5.5.9. 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.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during operating and accident conditions is 
restricted due to contact of the tube with the tubesheet. The 
leakage is modeled as flow through a porous medium through the use 
of the Darcy equation. The leakage model is used to develop a 
relationship between operational leakage and leakage at accident 
conditions that is based on differential pressure across the 
tubesheet and the viscosity of the fluid. A leak rate ratio was 
developed to relate the leakage at operating conditions to leakage 
at accident conditions. Since the fluid viscosity is based on fluid 
temperature and it is shown that for the most limiting accident, the 
fluid temperature does not exceed the normal operating temperature 
and therefore the viscosity ratio is assumed to be 1.0. Therefore, 
the leak rate ratio is a function of the ratio of the accident 
differential pressure and the normal operating differential 
pressure.
    The leakage factor of 1.93 for Braidwood Station Unit 2 and 
Byron Station Unit 2, for a postulated SLB/FLB, has been calculated 
as shown in Table 9-7 of WCAP-17072-P. However, EGC Braidwood 
Station Unit 2 and Byron Station Unit 2 will apply a factor of 3.11 
as determined by Westinghouse evaluation LTR-SGMP-09-100 P-
Attachment, Revision 1, to the normal operating leakage associated 
with the tubesheet expansion region in the condition monitoring (CM) 
and operational assessment (OA). The leakage factor of 3.11 applies 
specifically to Byron Unit 2 and Braidwood Unit 2, both hot and cold 
legs, in Table RAI24-2 of LTRSGMP-09-100 P-Attachment, Revision 1. 
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 assumed accident induced leak rate limit is 0.5 gallons 
per minute at room temperature (gpmRT) for the faulted SG and 0.218 
gpmRT for the unfaulted SGs for accidents that assume a faulted SG. 
These accidents are the SLB and the locked rotor with a stuck open 
PORV. The assumed accident induced leak rate limit for accidents 
that do not assume a faulted SG is 1.0 gpmRT for all SGs. These 
accidents are the locked rotor and control rod ejection.
    No leakage factor will be applied to the locked rotor or control 
rod ejection transients due to their short duration, since the 
calculated leak rate ratio is less than 1.0.
    The TS 3.4.13 operational leak rate limit is 150 gallons per day 
(gpd) (0.104 gpmRT) through any one SG. Consequently, there is 
sufficient margin between accident leakage and allowable operational 
leakage. The maximum accident leak rate ratio for the Model D5 
design SGs is 1.93 as indicated in WCAP-1 7072-P, Table 9-7. 
However, EGC will use the more conservative value of 3.11 accident 
leak rate ratio for the most limiting SG model design identified in 
Table RA124-2 of LTR-SGMP-09-100 P-Attachment Revision 1. This 
results in significant margin between the conservatively estimated 
accident leakage and the allowable accident leakage (0.5 gpmRT).
    For the CM assessment, the component of leakage from the prior 
cycle from below the H* distance will be multiplied by a factor of 
3.11 and added to the total leakage from any other source and 
compared to the allowable accident induced leakage limit. For the 
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 3.11 and compared to the 
observed operational leakage.
    Based on the above, the performance criteria of NEI-97-06, 
Revision 2, and draft RG 1.121 continue to be met and the proposed 
change does not involve a significant increase in the probability or 
consequences of the applicable accidents 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 change does not introduce any changes or mechanisms 
that create the possibility of a new or different kind of accident. 
Tube bundle integrity is expected to be maintained for all plant 
conditions upon implementation of the permanent alternate repair 
criteria. The proposed change does not introduce any new equipment 
or any change to existing equipment. No new effects on existing 
equipment are created nor are any new malfunctions introduced.
    Therefore, based on the above evaluation, 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?
    Response: No.
    The proposed change defines the safety significant portion of 
the SG tube that must be inspected and repaired. WCAP-17072-P as 
modified by WCAP-1 7330-P identifies the specific inspection depth 
below which any type tube degradation has no impact on the 
performance criteria in NEI 97-06, Revision 2, ``Steam Generator 
Program Guidelines.''
    The proposed change that alters the SG inspection and reporting 
criteria maintains the required structural margins of the SG tubes 
for both normal and accident conditions. NEI 97-06, and draft RG 
1.121 are used as the bases in the development of the limited 
tubesheet inspection depth methodology for determining that SG tube 
integrity considerations are maintained within acceptable limits. 
Draft 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. Draft 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 draft 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, WCAP-1 7072-P as modified by 
WCAP-17330-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-17072-P as modified by LTRSGMP-09-100 P-
Attachment shows that significant margin exists between an 
acceptable level of leakage during normal operating conditions that 
ensures meeting the SLB accident-induced leakage assumption and the 
TS leakage limit of 150 gpd.
    Based on the above, it is concluded that the proposed changes do 
not result in any 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: Robert D. Carlson.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station (CPS), Unit 1, DeWitt County, Illinois

    Date of amendment request: September 23, 2010.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would modify the CPS Technical Specifications (TS) Limiting

[[Page 5619]]

Condition for Operation (LCO) 3.7.6, ``Main Turbine Bypass System,'' by 
allowing revision of the reactor operational limits, as specified in 
the CPS Core Operating Limits Report (COLR), to compensate for the 
inoperability of the Main Turbine Bypass System (MTBS). The revised TS 
will require that either the MTBS be OPERABLE or that the reactor 
power, Minimum Critical Power Ratio (MCPR), and Linear Heat Generation 
Rate (LHGR) limits for an inoperable MTBS be placed in effect as 
specified in the COLR. Additionally, the amendment proposes modifying 
TS 5.6.5, ``Core Operating Limits Report (COLR),'' to add a requirement 
to establish cycle dependent reactor thermal power limits for an 
inoperable MTBS.
    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 MTBS functions to limit reactor pressure and power increases 
during certain transients postulated in the accident analysis. The 
MTBS is a mitigation function and not the initiator of any evaluated 
accident or transient. Operation with an inoperable MTBS while in 
compliance with the imposed reactor power limitation, and MCPR and 
LHGR limits will offset the impact of losing the MTBS function.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change will not create any new modes of plant or 
equipment operation. The proposed change allows the option to apply 
a reactor power penalty and an additional penalty factor to the MCPR 
and LHGR when the MTSS is inoperable. The imposed reactor power 
limitation and the revised set of MCPR and LHGR limits will offset 
the impact of losing the MTBS function, and maintain the margin to 
the MCPR safety limit and the thermal mechanical design limits.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    By establishing more restrictive reactor power and MCPR and LHGR 
operating limits, there are no changes to the plant design and 
safety analysis. There are no changes to the reactor core design 
instrument setpoints. The margin of safety assumed in the safety 
analysis is not affected. Applicable regulatory requirements will 
continue to be met and adequate defense-in-depth will be maintained. 
Sufficient safety margins will be maintained.
    The analytical methods used to determine the reactor power 
limitation and the revised core operating limits were reviewed and 
approved by the NRC and are described in Technical Specification 
5.6.5, ``Core Operating Limits Report (COLR).''
    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: Robert D. Carlson.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois

    Date of amendment request: October 4, 2010.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would revise Technical Specification (TS) Table 3.3.1.1 to 
eliminate Functions 5 and 10 from TS Table 3.3.1.1-1, delete footnote 
(c) from that table, and rename the footnote (d) to (c). These 
revisions would eliminate the requirement for a reactor scram, if 
vessel pressure is greater than or equal to 600 pounds per square inch 
gage (psig), with the reactor mode switch in startup and the main steam 
isolation valves closed or with a main turbine condenser vacuum low 
condition.
    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 proposed changes to the DNPS Units 2 and 3 TS revise the 
applicability of two protective functions and delete the associated 
TS Action statement. TS requirements that govern operability or 
routine testing of plant instruments are not assumed to be 
initiators of any analyzed event because these instruments are 
intended to prevent, detect, or mitigate accidents. Specifically, 
the reactor scram associated with the main steam isolation valve 
(MSIV) closure and low condenser vacuum (i.e., Functions 5 and 10 of 
TS 3.3.1.1) is in anticipation of the loss of the normal heat sink 
and subsequent overpressurization transient. The scram at high 
pressure in startup conditions when MSIVs close and/or main 
condenser vacuum is low does not impact the limiting accident or 
transient analyses. An analysis by General Electric Hitachi Nuclear 
Energy (GEH) demonstrated that the Mode 2 scram function for MSIV 
closure and low condenser vacuum can be eliminated without affecting 
safe plant operation. Elimination of these required scrams will not 
involve an increase in the probability of an accident previously 
evaluated.
    Additionally, these proposed changes will not increase the 
consequences of an accident previously evaluated because the 
proposed changes do not adversely impact structures, systems, or 
components. These changes will not alter the operation of equipment 
assumed to be available for the mitigation of accidents or 
transients by the plant safety analysis.
    Function 5 is currently required in Mode 2 with reactor pressure 
greater than or equal to 600 psig to ensure that the reactor is shut 
down, thus helping to prevent an overpressurization transient due to 
closure of main steam isolation valves. Similarly, Function 10 is 
currently required in Mode 2 with reactor pressure greater than or 
equal to 600 psig to help prevent an overpressurization transient by 
anticipating the turbine stop valve closure scram on loss of 
condenser vacuum.
    The existing scram logic is the result of experience gained 
during startup of an early vintage bailing water reactor in 1966 
when operators had difficulty controlling reactor power above 
approximately 600 psig without pressure control. Experience on later 
plant startups indicates that the early experience may not be 
inherent to later boiling water reactor designs. As such, GEH 
subsequently recommended elimination of the Mode 2 scram 
requirement.
    In Mode 2, the heat generation rate is low enough so that the 
other diverse Reactor Protection System (RPS) functions provide 
sufficient protection from an overpressurization transient. During 
normal power ascension in Mode 2 with the MSIVs open, reactor 
pressure vessel (RPV) pressure is controlled by the pressure 
regulator with increasing pressure setpoints. The maximum pressure 
regulator setpoint, which would translate to 1000 psig at rated 
power, would only allow a maximum dome pressure of approximately 900 
psig in the Mode 2 power range. The potential scenario in Mode 2 
whereby the MSIVs would close unexpectedly and cause the pressure to 
increase would lead to the Average Power Rate Monitors, Neutron 
Flux-High, Setdown scram (i.e., TS 3.3.1.1, Function 2.a), followed 
by the Reactor Vessel Steam Dome Pressure-High scram (i.e., TS 
3.3.1.1, Function 3).
    The consequences of a previously analyzed event are dependent on 
the initial conditions

[[Page 5620]]

assumed in the analysis, the availability and successful functioning 
of equipment assumed to operate in response to the analyzed event, 
and the setpoints at which these actions are initiated. The 
consequences of a previously evaluated accident are not 
significantly increased by the proposed change. The proposed change 
does not affect the performance of any equipment credited to 
mitigate the radiological consequences of an accident. Furthermore, 
there will be no change in the types or significant increase in the 
amounts of any effluents released offsite.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the DNPS Units 2 and 3 TS revise the 
applicability of two protective functions and delete the associated 
TS Action statement. The RPS functions are not an initiator of any 
accident. Rather, the RPS is designed to initiate a reactor scram 
when one or more monitored parameters exceed their specified limits 
to preserve the integrity of the fuel cladding and the reactor 
coolant pressure boundary and minimize the energy that must be 
absorbed following an accident. The proposed changes do not alter 
the applicability for RPS functions during plant conditions in which 
an overpressurization transient is assumed to occur. Specifically, 
no changes are being made to the required number of channels per 
trip system, surveillance requirements, or allowable values for 
these functions during Mode 1 operation.
    The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The proposed change does not change or 
introduce any new equipment, modes of system operation or failure 
mechanisms.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
and actions. The proposed changes revise the applicability for 
Functions 5 and 10 of TS 3.3.1.1 and delete an associated TS Action 
Statement. The proposed changes do not alter the applicability for 
RPS functions during plant conditions in which an overpressurization 
transient is assumed to occur.
    In addition, the proposed changes do not affect the probability 
of failure or availability of the affected instrumentation. 
Furthermore, the proposed changes will reduce the probability of 
test-induced plant transients and equipment failures.
    The proposed changes to the applicability for Functions 5 and 10 
of TS 3.3.1.1 have no impact on equipment design or fundamental 
operation. There are no changes being made to safety limits or 
safety system allowable values that would adversely affect plant 
safety. The performance of the systems important to safety is not 
significantly affected by the proposed changes. The proposed change 
does not affect safety analysis assumptions or initial conditions 
and therefore, the margin of safety in the original safety analyses 
is maintained.
    As documented above, 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: Robert. D. Carlson.

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: December 15, 2010.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
changes revise the Technical Specification (TS) relating to the Safety 
Limit Minimum Critical Power Ratios (SLMCPRs). The changes result from 
a cycle-specific analysis performed to support the operation of 
Limerick Generating Station, Unit 2, in the upcoming Cycle 12. 
Specifically, the proposed TS changes will revise the SLMCPRs contained 
in TS 2.1 for two recirculation loop operation and single recirculation 
loop operation to reflect the changes in the cycle-specific analysis. 
The new SLMCPRs are calculated using Nuclear Regulatory Commission 
(NRC)-approved methodology described in NEDE 24011-P-A, ``General 
Electric Standard Application for Reactor Fuel,'' Revision 17.
    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 derivation of the cycle specific Safety Limit Minimum 
Critical Power Ratios (SLMCPRs) for incorporation into the Technical 
Specifications (TS), and their use to determine cycle specific 
thermal limits, has been performed using the methodology discussed 
in NEDE-24011-P-A, ``General Electric Standard Application for 
Reactor Fuel,'' Revision 17.
    The basis of the SLMCPR calculation is to ensure that during 
normal operation and during abnormal operational transients, at 
least 99.9% of all fuel rods in the core do not experience 
transition boiling if the limit is not violated. The new SLMCPRs 
preserve the existing margin to transition boiling.
    The MCPR [minimum critical power ratio] safety limit is 
reevaluated for each reload using NRC-approved methodologies. The 
analyses for Limerick Generating Station (LGS), Unit 2, Cycle 12 
have concluded that a two loop MCPR safety limit of >=1.09, based on 
the application of Global Nuclear Fuel's 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 LGS, Unit 
2 Core Operating Limits Report (COLR).
    The requested TS 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 proposed TS changes do 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 SLMCPR is a TS numerical value, calculated to ensure that 
during normal operation and during abnormal operational transients, 
at least 99.9% of all fuel rods in the core do not experience 
transition boiling if the limit is not violated. The new SLMCPRs are 
calculated using NRC-approved methodology discussed in NEDE-24011-P-
A, ``General Electric Standard Application for Reactor Fuel,'' 
Revision 17. The proposed changes do not involve any new modes of 
operation or any plant modifications. The proposed revised MCPR 
safety limits have been shown to be acceptable for Cycle 12 
operation. The core operating limits will continue to be developed 
using NRC-approved methods. 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, the proposed TS changes do 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?

[[Page 5621]]

    Response: No.
    There is no significant reduction in the margin of safety 
previously approved by the NRC as a result of the proposed change to 
the SLMCPRs. The new SLMCPRs are calculated using methodology 
discussed in NEDE-24011-P-A, ``General Electric Standard Application 
for Reactor Fuel,'' Revision 17. The SLMCPRs ensure that during 
normal operation and during abnormal operational transients, at 
least 99.9% of all fuel rods in the core do not experience 
transition boiling if the limit is not violated, thereby preserving 
the fuel cladding integrity.
    Therefore, the proposed TS changes do not involve a significant 
reduction in the margin of safety previously approved by the NRC.

    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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: July 16, 2010, as supplemented by 
letters dated September 28, and November 23, 2010.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment to the Facility Operating License (FOL) includes: (1) The 
proposed Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS) Cyber 
Security Plan (the Plan), (2) an implementation schedule, and (3) 
revise the existing FOL Physical Protection license condition to 
require the FirstEnergy Nuclear Operating Company (FENOC, the licensee) 
to fully implement and maintain in effect all provisions of the 
Commission approved Cyber Security Plan as required by Title 10 of the 
Code of Federal Regulations (10 CFR) 73.54.
    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:

    Criterion 1: The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change is required by 10 CFR 73.54 and includes 
three parts. The first part is the submittal of the Plan for NRC 
review and approval. The Plan provides a description of how the 
requirements of the rule will be implemented at the DBNPS. The Plan 
establishes the licensing basis for the FENOC cyber security program 
for the DBNPS. The Plan establishes how to achieve high assurance 
that nuclear power plant digital computer and communication systems 
and networks associated with the following are adequately protected 
against cyber attacks up to and including the design basis threat:
    1. Safety-related and important-to-safety functions,
    2. Security functions,
    3. Emergency preparedness functions including offsite 
communications, and
    4. Support systems and equipment which if compromised, would 
adversely impact safety, security, or emergency preparedness 
functions.
    Part one of the proposed change is designed to achieve high 
assurance that the systems are protected from cyber attacks. The 
Plan itself does not require any plant modifications. However, the 
Plan does describe how plant modifications which involve digital 
computer systems are reviewed to provide high assurance of adequate 
protection against cyber attacks, up to and including the design 
basis threat as defined in the rule.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, affect the function of plant 
systems, or affect the manner in which systems are operated. The 
first part of the proposed change is designed to achieve high 
assurance that the systems within the scope of the rule are 
protected from cyber attacks and has no impact on the probability or 
consequences of an accident previously evaluated.
    The second part of the proposed change is an implementation 
schedule. The third part adds a sentence to the existing FOL license 
condition 2.D for Physical Protection. Both of these changes are 
administrative and have no impact on the probability or consequences 
of an accident previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2: The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change is required by 10 CFR 73.54 and includes 
three parts. The first part is the submittal of the Plan for NRC 
review and approval. The Plan provides a description of how the 
requirements of the rule will be implemented at the DBNPS. The Plan 
establishes the licensing basis for the FENOC cyber security program 
for the DBNPS. The Plan establishes how to achieve high assurance 
that nuclear power plant digital computer and communication systems 
and networks associated with the following are adequately protected 
against cyber attacks up to and including the design basis threat:
    1. Safety-related and important-to-safety functions,
    2. Security functions,
    3. Emergency preparedness functions including offsite 
communications, and
    4. Support systems and equipment which if compromised, would 
adversely impact safety, security, or emergency preparedness 
functions.
    Part one of the proposed change is designed to achieve high 
assurance that the systems within the scope of the rule are 
protected from cyber attacks. The Plan itself does not require any 
plant modifications. However, the Plan does describe how plant 
modifications which involve digital computer systems are reviewed to 
provide high assurance of adequate protection against cyber attacks, 
up to and including the design basis threat defined in the rule.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, affect the function of plant 
systems, or affect the manner in which systems are operated. The 
first part of the proposed change is designed to achieve high 
assurance that the systems within the scope of the rule are 
protected from cyber attacks and does not create the possibility of 
a new or different kind of accident from any previously evaluated.
    The second part of the proposed change is an implementation 
schedule. The third part adds a sentence to the existing FOL license 
condition 2.D for Physical Protection. Both of these changes are 
administrative and do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Criterion 3: The proposed change does not involve a significant 
reduction in a margin of safety.
    The proposed change is required by 10 CFR 73.54 and includes 
three parts. The first part is the submittal of the Plan for NRC 
review and approval. The Plan provides a description of how the 
requirements of the rule will be implemented at the DBNPS. The Plan 
establishes the licensing basis for the FENOC cyber security program 
for the DBNPS. The Plan establishes how to achieve high assurance 
that nuclear power plant digital computer and communication systems 
and networks associated with the following are adequately protected 
against cyber attacks up to and including the design basis threat:
    1. Safety-related and important-to-safety functions,
    2. Security functions,
    3. Emergency preparedness functions including offsite 
communications, and
    4. Support systems and equipment which if compromised, would 
adversely impact safety, security, or emergency preparedness 
functions.
    Part one of the proposed change is designed to achieve high 
assurance that the systems within the scope of the rule are 
protected from cyber attacks. Plant safety margins are established 
through Limiting Conditions for Operation, Limiting Safety System 
Settings and Safety limits specified in

[[Page 5622]]

the Technical Specifications, methods of evaluation that establish 
design basis or change Updated Final Safety Analysis. Because there 
is no change to these established safety margins, the proposed 
change does not involve a significant reduction in a margin of 
safety.
    The second part of the proposed change is an implementation 
schedule. The third part adds a sentence to the existing FOL license 
condition 2.D for Physical Protection. Both of these changes are 
administrative and do 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: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: Robert. D. Carlson.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, 
Texas

    Date of amendment request: December 1, 2010.
    Brief description of amendments: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would revise Technical Specification (TS) 5.5.9, ``Unit 1 
Model D76 and Unit 2 Model D5 Steam Generator (SG) Program,'' to 
exclude portions of the Unit 2 Model D5 steam generator (SG) tubes 
below the top of the SG tubesheet from periodic SG tube inspections 
during Comanche Peak Nuclear Power Plant (CPNPP), Unit 2 Refueling 
Outage 12 and the subsequent operating cycle. In addition, the proposed 
amendment would revise TS 5.6.9, ``Unit 1 Model D76 and Unit 2 Model D5 
Steam Generator Tube Inspection Report,'' to provide reporting 
requirements specific to CPNPP, Unit 2 for the temporary alternate 
repair criteria.
    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.
    Of the 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.
    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, differential pressure 
between the primary and secondary side, and tubesheet rotation. 
Based on this design, the structural margins against burst, as 
discussed in [NRC] Regulatory Guide (RG) 1.121, ``Bases for Plugging 
Degraded PWR [Pressurized-Water Reactor] Steam Generator Tubes,'' 
and TS 5.5.9 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 in TS 5.5.9. 
Therefore, the proposed change results in no significant increase in 
the probability of the occurrence of a[n] 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[n] SGTR.
    The probability of a[n] SLB is unaffected by the potential 
failure of a steam generator tube as the failure of tube is not an 
initiator for a[n] SLB event.
    The leakage factor of 3.16 for CPNPP Unit 2, for a postulated 
SLB/FLB, has been calculated as described in Reference 8.29 
[Westinghouse Letter LTR-SGMP-09-100P-Attachment, Revision 1, dated 
September 7, 2010] and is shown in Revised Table 9-7 of this same 
reference. Specifically, 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 3.16 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 3.16 and compared to 
the observed operational leakage. The accident-induced leak rate 
limit for CPNPP Unit 2 is 1.0 gpm [gallons per minute]. The TS 
operational leak rate limit through any one steam generator is 150 
gpd [gallons per day] (0.1 gpm). Consequently, there is significant 
margin between accident leakage and allowable operational leakage. 
The SLB/FLB overall leakage factor is 3.16 resulting in significant 
margin between the conservatively estimated accident induced leakage 
and the allowable accident leakage.
    No leakage factor was applied to the locked rotor or control rod 
ejection transients due to their short duration.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the 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.
    Therefore, 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 previously evaluated?
    Response: No.
    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. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    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, Rev. 2, ``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 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[n] SGTR. RG 1.121 
concludes that by determining the limiting safe conditions for tube 
wall degradation, the probability and

[[Page 5623]]

consequences of a[n] SGTR are reduced. RG 1.121 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) 
[Boiler and Pressure Vessel] Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, the H* Analysis documented in 
Section 4.1 [Attachment 1 to letter dated December 1, 2010] 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 provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: November 30, 2010.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would revise the Wolf Creek Generating Station's (WCGS's) 
Technical Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' 
to exclude portions of the tube below the top of the steam generator 
tubesheet from periodic steam generator tube inspections during 
Refueling Outage 18 and the subsequent operating cycle. In addition, 
the proposed amendment would revise TS 5.6.10, ``Steam Generator Tube 
Inspection Report,'' to remove references to previous interim alternate 
repair criteria and provide reporting requirements specific to the 
temporary alternate repair criteria.
    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 previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the steam generator inspection 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 steam 
generator tube inspection and repair criteria are the steam 
generator tube rupture (SGTR) event and the feedline break (FLB) 
postulated accidents.
    During the SGTR event, the required structural integrity margins 
of the steam generator 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 presence of the 
tubesheet and constraint provided by 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, from the differential pressure between the primary and 
secondary side, and tubesheet deflection. 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 TS 5.5.9 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 
steam generator tubes consistent with the performance criteria in TS 
5.5.9. Therefore, the proposed change results in no significant 
increase in the probability of the occurrence of a[n] SGTR accident.
    At normal operating pressures, leakage from tube degradation 
below the proposed limited inspection depth is limited by the tube-
to-tubesheet joint. 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 changes do not result in 
a significant increase in the consequences of a[n] SGTR.
    The consequences of a steam line break (SLB) are also not 
significantly affected by the proposed changes. During a[n] SLB 
accident, the reduction in pressure above the tubesheet on the shell 
side of the steam generator creates an axially uniformly distributed 
load on the tubesheet due to the reactor coolant system pressure on 
the underside of the tubesheet. The resulting bending action 
constrains the tubes in the tubesheet thereby restricting primary-
to-secondary leakage below the midplane.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., an SLB) is 
limited by flow restrictions. These restrictions result from the 
crack and tube-to-tubesheet contact pressures that provide a 
restricted leakage path above the indications and also limit the 
degree of potential crack face opening as compared to free span 
indications.
    The leakage factor of 2.50 for WCGS, for a postulated SLB/FLB, 
has been calculated as shown in Revised Table 9-7 of Reference 15 
[Westinghouse Letter LTR-SGMP-09-100, dated August 12, 2009]. 
Specifically, 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 2.50 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 2.50 and compared to the 
observed operational leakage.
    The probability of an SLB is unaffected by the potential failure 
of a steam generator tube as the failure of the tube is not an 
initiator for an SLB event. SLB leakage is limited by leakage flow 
restrictions resulting from the leakage path above potential cracks 
through the tube-to-tubesheet crevice. The leak rate during 
postulated accident conditions (including locked rotor) has been 
shown to remain within the accident analysis assumptions for all 
axial and or circumferentially orientated cracks occurring 15.2 
inches below the top of the tubesheet. The accident induced leak 
rate limit for WCGS is 1.0 gpm [gallon per minute]. The TS 3.4.13, 
``RCS [Reactor Coolant System] Operational LEAKAGE,'' operational 
leak rate limit is 150 gpd [gallons per day] (0.1 gpm) through 
anyone steam generator. Consequently, accident leakage is 
approximately 10 times the allowable leakage, if only one steam 
generator is leaking. Using an SLB/FLB overall leakage factor of 
2.50, accident induced leakage is approximately 0.5 gpm, if all 4 
steam generators are leaking at 150 gpd at the beginning of the 
accident. Therefore, significant margin exists between the 
conservatively estimated accident induced leakage and the allowable 
accident leakage (1.0 gpm).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 5624]]

    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change alters the steam generator inspection and 
reporting criteria. It 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 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 accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed change alters the steam generator inspection and 
reporting criteria. It maintains the required structural margins of 
the steam generator tubes for both normal and accident conditions. 
NEI [Nuclear Energy Institute] 97-06, Revision 2, and RG 1.121, are 
used as the bases in the development of the limited tubesheet 
inspection depth methodology for determining that steam generator 
tube integrity considerations are maintained within acceptable 
limits. RG 1.121 describes a method acceptable to the NRC for 
meeting GDC [General Design Criterion] 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[n] SGTR. RG 1.121 
concludes that by determining the limiting safe conditions for tube 
wall degradation, the probability and consequences of a[n] 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) [Boiler and Pressure Vessel] 
Code. For axially-oriented cracking located within the tubesheet, 
tube burst is precluded due to the presence of the tubesheet. For 
circumferentially-oriented cracking, the H* Analysis documented in 
Section 3 [of letter dated November 30, 2010], 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 provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Unit 2 and 3, New London County, 
Connecticut
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station (CPS), Unit 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania
Exelon Generation Company, LLC
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, 
Texas
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    A. This Order contains instructions regarding how potential parties 
to this proceeding may request access to documents containing Sensitive 
Unclassified Non-Safeguards Information (SUNSI).
    B. Within 10 days after publication of this notice of hearing and 
opportunity to petition for leave to intervene, any potential party who 
believes access to SUNSI is necessary to respond to this notice may 
request such access. A ``potential party'' is any person who intends to 
participate as a party by demonstrating standing and filing an 
admissible contention under 10 CFR 2.309. Requests for access to SUNSI 
submitted later than 10 days after publication will not be considered 
absent a showing of good cause for the late filing, addressing why the 
request could not have been filed earlier.
    C. The requestor shall submit a letter requesting permission to 
access SUNSI to the Office of the Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, and provide a copy to the Associate General 
Counsel for Hearings, Enforcement and Administration, Office of the 
General Counsel, Washington, DC 20555-0001. The expedited delivery or 
courier mail address for both offices is: U.S. Nuclear Regulatory 
Commission, 11555 Rockville Pike, Rockville, Maryland 20852. The e-mail 
address for the Office of the Secretary and the Office of the General 
Counsel are <a href="/cdn-cgi/l/email-protection" class="__cf_email__" data-cfemail="fcb4999d8e95929bd2b8939f979988bc928e9fd29b938a">[email protected]</a> and <a href="/cdn-cgi/l/email-protection#2c636b6f414d45404f494258495e6c425e4f024b435a"><span class="__cf_email__" data-cfemail="460901052b272f2a2523283223340628342568212930">[email protected]</span></a>, 
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 
those documents. These conditions may include, but are not limited to, 
the signing of a Non-Disclosure Agreement

[[Page 5625]]

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 Grants of Access. A party other than the requestor may 
challenge an NRC staff determination granting access to SUNSI whose 
release would harm that party's interest independent of the proceeding. 
Such a challenge must be filed with the Chief Administrative Judge 
within 5 days of the notification by the NRC staff of its grant of 
access.
    If challenges to the NRC staff determinations are filed, these 
procedures give way to the normal process for litigating disputes 
concerning access to information. The availability of interlocutory 
review by the Commission of orders ruling on such NRC staff 
determinations (whether granting or denying access) is governed by 10 
CFR 2.311.\3\
---------------------------------------------------------------------------

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

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

    Dated at Rockville, Maryland, this 25th day of January 2011.
    For the Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.

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

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

[[Page 5626]]

 
>A + 60.......................  Decision on contention admission.
------------------------------------------------------------------------


[FR Doc. 2011-2027 Filed 1-26-11; 4:15 pm]
BILLING CODE 7590-01-P


</pre><script data-cfasync="false" src="/cdn-cgi/scripts/5c5dd728/cloudflare-static/email-decode.min.js"></script></body>
</html>