[Federal Register Volume 75, Number 2 (Tuesday, January 5, 2010)]
[Notices]
[Pages 458-467]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-31060]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0564]
Notice; Applications and Amendments to Facility Operating
Licenses Involving Proposed No Significant Hazards Considerations and
Containing Sensitive Unclassified Non-Safeguards Information and Order
Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards
Information
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this notice. The Act requires
the Commission publish notice of any amendments issued, or proposed to
be issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license upon a
determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This notice includes notices of amendments containing sensitive
unclassified non-safeguards information (SUNSI).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking
and Directives Branch (RDB), TWB-05-
[[Page 459]]
B01M, Division of Administrative Services, Office of Administration,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and
should cite the publication date and page number of this Federal
Register notice. Written comments may also be faxed to the RDB at 301-
492-3446. Documents may be examined, and/or copied for a fee, at the
NRC's Public Document Room (PDR), located at One White Flint North,
Public File Area O1 F21, 11555 Rockville Pike (first floor), Rockville,
Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, or
at 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, 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
[email protected], or by telephone at (301) 415-1677, to request
(1) a digital ID certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on NRC's public Web site at 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 EIE, users will be required to install a Web
browser plug-
[[Page 460]]
in from the NRC Web site. Further information on the Web-based
submission form, including the installation of the Web browser plug-in,
is available on the NRC's public Web site at 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
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
social security numbers, home addresses, or home phone numbers in their
filings, unless an NRC regulation or other law requires submission of
such information. With respect to copyrighted works, except for limited
excerpts that serve the purpose of the adjudicatory filings and would
constitute a Fair Use application, participants are requested not to
include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from January 5, 2010. Non-timely filings will not be entertained
absent a determination by the presiding officer that the petition or
request should be granted or the contentions should be admitted, based
on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-
(viii).
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible electronically from the
ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, 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 [email protected].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant (HBRSEP) Unit No. 2, Darlington County, South Carolina
Date of amendment request: June 19, 2009, as supplemented by letter
dated October 20, 2009.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendment would revise TS 3.3.1, ``Reactor Protection System
Instrumentation.'' The proposed change revises the requirements related
to the reactor protection system interlock for the turbine trip input
to the reactor protection system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change provides revised requirements for the
reactor protection system interlock associated with the turbine trip
protection function. The proposed change will allow the interlock
for turbine trip function to be raised from the current interlock
setting of nominally 10 percent reactor power to nominally 40
percent reactor power.
This change will allow the reactor to continue operating safely
at power levels up to nominally 40 percent when the turbine is not
operating. The applicable accident analyses, as described in the
HBRSEP, Unit No. 2, Updated Final Safety Analysis Report (UFSAR)
have been reviewed. The turbine trip input to reactor trip has been
verified to be either not used in the accident analyses or that the
change does not adversely affect the analyses results and
conclusion. Therefore, it is concluded that the consequences as
described in the UFSAR accident analyses are unaffected by the
proposed change.
An analysis of plant response to a turbine trip at nominally 40
percent power provided with the amendment request shows that the
applicable acceptance criteria are met. Specifically, analysis has
shown that a turbine trip without a reactor trip below 40 percent
power does not challenge the pressurizer PORVs [power operated
relief valves] or the steam generator safety valves; thereby, not
adversely affecting the probability of a small break LOCA [loss of
coolant accident] due to a stuck open PORV, or an excessive cooldown
event due to a stuck open steam generator safety valve. As a result,
the probability of any accident
[[Page 461]]
previously evaluated is not significantly increased by the proposed
changes.
Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated.
As described above, the proposed change provides revised
requirements for the reactor protection system interlock associated
with the turbine trip protection function. The proposed change will
allow the interlock for turbine trip function to be raised from the
current interlock setting of nominally 10 percent reactor power to
nominally 40 percent reactor power.
No new accident initiators or precursors are introduced by the
proposed change. Changing the interlock for the reactor trip on
turbine trip from P-7 to P-8 changes the power level associated with
enabling and disabling the reactor trip on turbine trip function.
The turbine pressure input to the reactor protection system
permissive is not an accident initiator. The change does not affect
how the associated trip functional units operate or function. The
changes do not create the possibility of a new or different kind of
accident from any previously evaluated because these interlock
changes do not affect the way that the associated trip functional
units operate or function.
Therefore, operation of the facility in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
As described above, the proposed change provides revised
requirements for the reactor protection system interlock associated
with the turbine trip protection function. The proposed change will
allow the interlock for the turbine trip function to be raised from
the current interlock setting of nominally 10 percent reactor power
to nominally 40 percent reactor power.
Also, as previously described, this change will allow the
reactor to continue operating safely at power levels up to nominally
40 percent when the turbine is not operating. The applicable UFSAR
accident analyses have been reviewed and it is concluded that the
accident analyses are unaffected by the proposed change. An analysis
of plant response to a turbine trip at nominally 40 percent power
shows that the applicable acceptance criteria are met. Based on
these evaluations, the margins of safety that could potentially have
been impacted by the proposed change associated with the reactor,
which include departure from nucleate boiling (DNB) and fuel
temperature margins, and the margin of safety associated with
reactor coolant system integrity, are not affected.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: October 27, 2009.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). This
amendment request would change the Technical Specifications to provide
revised values for the Safety Limit Minimum Critical Power Ratio
(SLMCPR) for both single and dual recirculation loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The basis of the Safety Limit MCPR (SLMCPR) is to ensure no
mechanistic fuel damage is calculated to occur if the limit is not
violated. The new SLMCPR values preserve the existing margin to
transition boiling and probability of fuel damage is not increased.
The derivation of the revised SLMCPR for Vermont Yankee for
incorporation into the Technical Specifications and its use to
determine plant and cycle-specific thermal limits has been performed
using NRC approved methods. These plant-specific calculations are
performed each operating cycle and if necessary, will require future
changes to these values based upon revised core designs. The revised
SLMCPR values do not change the method of operating the plant and
have no effect on the probability of an accident initiating event or
transient.
Based on the above, Vermont Yankee has concluded that the
proposed change will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes result only from a specific analysis for
the Vermont Yankee core reload design. These changes do not involve
any new or different methods for operating the facility. No new
initiating events or transients result from these changes.
Based on the above, Vermont Yankee has concluded that the
proposed change will not create the possibility of a new or
different kind of accident from those previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The new SLMCPR is calculated using NRC approved methods with
plant and cycle specific parameters for the current core design. The
SLMCPR value remains conservative enough to ensure that greater than
99.9% of all fuel rods in the core will avoid transition boiling if
the limit is not violated, thereby preserving the fuel cladding
integrity. The operating MCPR limit is set appropriately above the
safety limit value to ensure adequate margin when the cycle specific
transients are evaluated. Accordingly, the margin of safety is
maintained with the revised values.
As a result, Vermont Yankee has determined that the proposed
change will not result in a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy L. Salgado.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: October 27, 2009.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendment revises the Technical Specifications to increase the two
recirculation loop minimum critical power ratio (MCPR) safety limit
from 1.08 to 1.09 and the single recirculation loop MCPR safety limit
from 1.10 to 1.12.
[[Page 462]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Minimum Critical Power Ratio (MCPR) limit is defined in the
Bases to Technical Specification 2.1.1.2 as that limit, ``that, in
the event of an AOO [(Anticipated Operational Occurrence)] from the
limiting condition of operation, at least 99.9% of the fuel rods in
the core would be expected to avoid boiling transition.'' The MCPR
safety limit satisfies the requirements of General Design Criterion
10 of Appendix A to 10CFR50 regarding acceptable fuel design limits.
The MCPR safety limit is reevaluated for each reload using NRC-
approved methodologies. The analyses for GGNS [Grand Gulf Nuclear
Station] Cycle 18 have concluded that a two-loop MCPR safety limit
of 1.09, based on the application of Global Nuclear Fuels' NRC
approved MCPR safety limit methodology, will ensure that this
acceptance criterion is met. For single-loop operation, a MCPR
safety limit of 1.12, also ensures that this acceptance criterion is
met. The MCPR operating limits are presented and controlled in
accordance with the GGNS Core Operating Limits Report (COLR).
The requested Technical Specification changes do not involve any
plant modifications or operational changes that could affect system
reliability or performance or that could affect the probability of
operator error. The requested changes do not affect any postulated
accident precursors, do not affect any accident mitigating systems,
and do not introduce any new accident initiation mechanisms.
Therefore, the changes to the Minimum Critical Power Ratio
safety limit do not involve a significant increase in the
probability or consequences of any accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The GNF2 fuel to be used in Cycle 18 is of a design compatible
with the co-resident GE14 and ATRIUM-10 fuel. Therefore, the
introduction of GNF2 fuel into the Cycle 18 core will not create the
possibility of a new or different kind of accident. The proposed
changes do not involve any new modes of operation, any changes to
setpoints, or any plant modifications. The proposed revised MCPR
safety limits have accounted for the mixed fuel core and have been
shown to be acceptable for Cycle 18 operation. Compliance with the
criterion for incipient boiling transition continues to be ensured.
The core operating limits will continue to be developed using NRC
approved methods which also account for the mixed fuel core design.
The proposed MCPR safety limits or methods for establishing the core
operating limits do not result in the creation of any new precursors
to an accident.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The MCPR safety limits have been evaluated in accordance with
Global Nuclear Fuels NRC-approved cycle-specific safety limit
methodology to ensure that during normal operation and during AOO's
at least 99.9% of the fuel rods in the core are not expected to
experience transition boiling. The proposed revised MCPR safety
limits have accounted for the mixed fuel core and have been shown to
be acceptable for Cycle 18 operation. Compliance with the criterion
for incipient boiling transition continues to be ensured. On this
basis, the implementation of the change to the MCPR safety limits
does not involve a significant reduction in a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: November 3, 2009.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendment revises the Technical Specifications (TSs) to reflect the
installation of the digital General Electric--Hitachi Nuclear
Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring
(PRNM) System.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability (frequency of occurrence) of design basis
accidents (DBAs) occurring is not affected by the NUMAC PRNM System,
since the system does not interact with equipment whose failure
could cause an accident. Compliance with the regulatory criteria
established for plant equipment are maintained with the installation
of the upgraded NUMAC PRNM System. Scram setpoints in the NUMAC PRNM
System are established such that the analytical limits are met.
The unavailability of the new NUMAC PRNM System is equal to or
less than the existing system and, as a result, the scram
reliability is equal to or better than the existing analog power
system. No new challenges to safety-related equipment result from
the NUMAC PRNM System modification. Therefore, the proposed change
does not involve a significant increase in the probability of an
accident previously evaluated.
The proposed change replaces the current Option E-I-A stability
solution with an NRC-approved Option III long-term stability
solution. The NUMAC PRNM hardware incorporates the Oscillation Power
Range Monitor (OPRM) Option III detect-and-suppress solution, which
has been previously reviewed and approved by the NRC. The OPRM meets
[10 CFR Part 50, Appendix A] General Design Criterion (GDC) 10,
Reactor Design, and GDC 12, Suppression of Reactor Power
Oscillations, requirements by automatically detecting and
suppressing design basis thermal-hydraulic oscillations prior to
exceeding the fuel Minimum Critical Power Ratio (MCPR) Safety Limit.
Based on the above, installation of the new NUMAC PRNM System
with the OPRM Option III stability solution integrated into the
NUMAC PRNM equipment does not increase the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The components of the NUMAC PRNM System are equivalent or of
better design and qualification criteria than those currently
installed and utilized in the plant. No new operating mode, safety-
related equipment lineup, accident scenario, or interaction mode not
reviewed and approved as part of the design and licensing of the
NUMAC PRNM System has been identified. Therefore, the NUMAC PRNM
System retrofit does not adversely affect plant equipment.
The new NUMAC PRNM System uses digital equipment that has
software-controlled digital processing compared to the existing
power range system that uses mostly analog and discrete component
processing. Specific failures of hardware and potential software
common-cause failures are different from the existing system. The
effects of potential software common-cause failure are mitigated by
specific hardware design and
[[Page 463]]
system architecture as discussed in Section 6.0 of [GE Nuclear
Energy Licensing Topical Report] NEDC-32410P-A. Failure(s) of the
system have the same overall effect as the present design. No new or
different kinds of accidents are introduced. Therefore, the NUMAC
PRNM System does not adversely effect plant equipment.
The currently installed Average Power Range Monitoring (APRM)
system is replaced with a NUMAC PRNM System that performs the
existing power range monitoring functions and adds an OPRM to react
automatically to potential reactor thermal-hydraulic instabilities.
Based on the above, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed TS changes associated with the NUMAC PRNM System
retrofit implement the constraints of the NUMAC PRNM System design
and related stability analyses. The NUMAC PRNM System change does
not impact reactor operating parameters or the functional
requirements of the APRM system. The replacement equipment continues
to provide information, enforce control rod blocks, and initiate
reactor scrams under appropriate specified conditions. The proposed
change does not reduce safety margins. The replacement APRM
equipment has improved channel trip accuracy compared to the current
analog system, and meets or exceeds system requirements previously
assumed in setpoint analysis. Thus, the ability of the new equipment
to enforce compliance with margins of safety equals or exceeds the
ability of the equipment which it replaces.
Therefore, the proposed changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: October 5, 2009.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The
amendment(s) would revise Technical Specification (TS) 4.3.1,
``Criticality,'' to address a non-conservative TS. The proposed change
addresses the Boraflex degradation issue in the LaSalle County Station
(LSCS) Unit 2 spent fuel storage racks by revising TS Section 4.3.1 to
allow the use of NETCO-SNAP-IN[reg] rack inserts in LSCS Unit 2 spent
fuel storage rack cells as a replacement for the neutron absorbing
properties of the existing Boraflex panels.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change adds an additional requirement to TS Section
4.3.1 to install a NETCO-SNAP-IN[reg] rack insert in spent fuel
storage rack cells that cannot otherwise maintain the requirements
of TS Section 4.3.1.1.a to ensure that the effective neutron
multiplication factor, Keff, is less than or equal to
0.95, if the spent fuel pool (SFP) is fully flooded with unborated
water. The proposed change also includes a revision to TS Section
4.3.1 to specify the bounding reactivity fuel design allowed for
storage in the Unit 1 and Unit 2 SFPs. Since the proposed change
pertains only to the SFP, only those accidents that are related to
movement and storage of fuel assemblies in the SFP could be
potentially affected by the proposed change.
The current licensing basis for the LSCS Unit 2 SFP credits the
neutron absorbing properties of the Boraflex neutron poison material
in the spent fuel storage racks. The current licensing basis
demonstrates: (1) Adequate margin to criticality for spent fuel
storage rack cells that credit the neutron absorption capabilities
of Boraflex, (2) adequate margin for fuel assemblies inadvertently
placed into locations adjacent to the spent fuel storage racks, and
(3) adequate margin for assemblies accidentally dropped onto the
spent fuel storage racks. Therefore, the probability that a
misplaced fuel assembly would result in an inadvertent criticality
is unchanged since the process and procedural controls governing
fuel movement in the SFP will not be changed. The dose consequences
of the most limiting drop of a fuel assembly in the SFP is limited
by the number of the fuel rods damaged and other engineered features
unaffected by the proposed change, including the fuel design, fuel
decay time, water level in the SFP, water temperature of the SFP,
and the engineering features of the Reactor Building Ventilation
System.
The installation of NETCO-SNAP-IN[reg] rack inserts does not
result in a significant increase in the probability of an accident
previously analyzed. The revised criticality analysis takes no
credit for the Boraflex material. The use of a rack insert provides
an alternative neutron absorber to take the place of the degraded
Boraflex material, without removal of the existing Boraflex. The
probability that a fuel assembly would be dropped is unchanged by
the installation of the NETCO-SNAP-IN[reg] rack inserts. These
events involve failures of administrative controls, human
performance, and equipment failures that are unaffected by the
presence or absence of Boraflex and the rack inserts.
The installation of NETCO-SNAP-IN[reg] rack inserts does not
result in a significant increase in the consequence of an accident
previously analyzed. A criticality analysis has been prepared to
demonstrate adequate margin to criticality for spent fuel storage
rack cells with rack inserts in the LSCS Unit 2 SFP, and adequate
criticality margin for assemblies accidentally dropped onto the
spent fuel storage racks.
The installation of NETCO-SNAP-IN@ rack inserts does not affect
the consequences of a dropped fuel assembly. The consequences of
dropping a fuel assembly onto any other fuel assembly or other
structure are unaffected by the change. The consequences of dropping
a fuel assembly onto a spent fuel storage rack cell with a rack
insert are bounded by the event of dropping an assembly onto another
assembly, both for criticality and for radiological consequences.
For criticality, the effects on Keff of dropping a fuel
assembly have been evaluated and are acceptable. For radiological
consequences, the number of rods damaged when a fuel assembly is
accidentally dropped onto a spent fuel storage rack cell with or
without a rack insert is bounded by the number of rods damaged by an
assembly dropped onto another assembly. The change does not affect
the effectiveness of the other engineered design features to limit
the offsite dose consequences of the limiting fuel assembly drop
accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Onsite storage of spent fuel assemblies in the SFP is a normal
activity for which LSCS has been designed and licensed. As part of
assuring that this normal activity can be performed without
endangering public health and safety, the ability to safely
accommodate different possible accidents in the SFP, such as
dropping a fuel assembly or misloading a fuel assembly, have been
analyzed. The proposed spent fuel storage configuration does not
change the methods of fuel movement or spent fuel storage. The
proposed change allows for continued use of spent fuel storage rack
cells that have been determined unusable based on the degradation of
Boraflex within those spent fuel storage rack cells. The rack
inserts are passive devices. These devices, when inside a spent fuel
storage rack cell, perform the same function as the Boraflex in that
cell without the potential for degradation. These devices do not add
any limiting structural loads or affect the removal of decay heat
from
[[Page 464]]
the assemblies. No change in total heat load in the SFP is being
made. The devices are resistant to corrosion and will maintain their
structural integrity over the life of the SFP. An accidental fuel
assembly drop does not challenge their structural integrity. The
existing fuel handling accident, which assumes the drop of a fuel
assembly, bounds the drop of a rack insert and/or rack insert
installation tool. This change does not create the possibility of a
misloaded assembly into a spent fuel storage rack cell.
The misloading of a more reactive assembly targeted for
placement in the LSCS Unit 1 SFP or the LSCS Unit 2 SFP Boraflex
region in a rack insert region of the LSCS Unit 2 SFP has been
prevented since the most reactive fuel assembly at LSCS is bounded
by the rack insert criticality analysis, and the most reactive fuel
assembly allowed for future insertion in either the Unit 1 or Unit 2
SFP is being limited to the reference bounding ATRIUM-10 fuel
assembly.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
LSCS TS 4.3.1.1 requires the spent fuel storage racks to
maintain the effective neutron multiplication factor,
Keff, less than or equal to 0.95 when fully flooded with
unborated water, which includes an allowance for uncertainties.
Therefore, for criticality, the required safety margin is 5%
including a conservative margin to account for engineering and
manufacturing uncertainties.
The proposed change provides an alternative method to ensure
that Keff continues to be less than or equal to 0.95,
thus preserving the required safety margin of 5%. The criticality
analysis demonstrates the required margin to criticality of 5%,
including a conservative margin to account for engineering and
manufacturing uncertainties, is maintained assuming an infinite
array of fuel with all fuel at the peak reactivity. In addition, the
margin of safety for radiological consequences of a dropped fuel
assembly are unchanged because the event involving a dropped fuel
assembly onto a spent fuel storage rack cell containing a fuel
assembly with a rack insert is bounded by the consequences of a
dropped fuel assembly without a rack insert. The proposed change
also maintains the capacity of the Unit 2 SFP to be no more than
4078 fuel assemblies.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station,
Unit No. 1, Salem County, New Jersey
Date of amendment request: October 8, 2009.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendment would revise Technical Specification (TS) 6.8.4.i, ``Steam
Generator (SG) Program,'' by adding a one-time alternate repair
criterion that excludes certain portions of the tube below the top of
the SG tubesheet from periodic SG tube inspections. In addition, the
proposed amendment would revise TS 6.9.10, ``Steam Generator Tube
Inspection Report,'' to provide reporting requirements specific to the
alternate repair criteria. The proposed amendment is supported by
Westinghouse Electric Company, LLC Topical Report WCAP-17071-P, ``H*:
Alternate Repair Criteria for the Tubesheet Expansion Region in Steam
Generators with Hydraulically Expanded Tubes (Model F).'' H*
(pronounced ``H star'') is the length of hydraulically expanded SG tube
that must remain intact within the tubesheet in order for the joint to
resist pullout and leakage due to normal operating and accident
conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
that alters the steam generator (SG) inspection and reporting
criteria does not have a detrimental impact on the integrity of any
plant structure, system, or component that initiates an analyzed
event. The proposed change will not alter the operation of, or
otherwise increase the failure probability of any plant equipment
that initiates an analyzed accident.
Of the applicable accidents previously evaluated, the limiting
transients with consideration to the proposed change to the SG tube
inspection and repair criteria are the steam generator tube rupture
(SGTR) event, the steam line break (SLB), and the feed line break
(FLB) postulated accidents.
During the SGTR event, the required structural integrity margins
of the SG tubes and the tube-to-tubesheet joint over the H* distance
will be maintained. Tube rupture in tubes with cracks within the
tubesheet is precluded by the constraint provided by the presence of
the tubesheet and the tube-to-tubesheet joint. Tube burst cannot
occur within the thickness of the tubesheet. The tube-to-tubesheet
joint constraint results from the hydraulic expansion process,
thermal expansion mismatch between the tube and tubesheet, and from
the differential pressure between the primary and secondary side,
and tubesheet rotation. Based on this design, the structural margins
against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases
for Plugging Degraded PWR [pressurized-water reactor] Steam
Generator Tubes,'' and Technical Specification 6.8.4.i, are
maintained for both normal and postulated accident conditions.
The proposed change has no impact on the structural or leakage
integrity of the portion of the tube outside of the tubesheet. The
proposed change maintains structural and leakage integrity of the SG
tubes consistent with the performance criteria of Technical
Specification 6.8.4.i. Therefore, the proposed change results in no
significant increase in the probability of the occurrence of a SGTR
accident.
At normal operating pressures, leakage from tube degradation
below the proposed limited inspection depth is limited by the tube-
to-tubesheet crevice. Consequently, negligible normal operating
leakage is expected from degradation below the inspected depth
within the tubesheet region. The consequences of an SGTR event are
not affected by the primary-to-secondary leakage flow during the
event as primary-to-secondary leakage flow through a postulated tube
that has been pulled out of the tubesheet is essentially equivalent
to a severed tube. Therefore, the proposed change does not result in
a significant increase in the consequences of a SGTR[.]
The probability of a SLB is unaffected by the potential failure
of a steam generator tube as the failure of tube is not an initiator
for a SLB event.
The leakage factor of 2.16 for Salem Unit 1, for a postulated
SLB/FLB, has been calculated as shown in Table 9-7 of WCAP-17071-P
as revised by the response to RAI [request for additional
information] 24 (Attachment 7 [to the application dated October 8,
2009]). Through application of the limited tubesheet inspection
scope, the existing operating leakage limit provides assurance that
excessive leakage (i.e., greater than accident analysis assumptions)
will not occur. The accident analysis calculations have an
assumption of 0.6 [gallons per minute (gpm)] at room temperature
(gpmRT) primary-to-secondary leakage in a single SG and 1 gpm at
room temperature (gpmRT) total primary-to-secondary leakage for all
SGs. This apportioned primary-to-secondary leakage is used in the
Main Steam Line Break and Locked Rotor accidents. Primary-to-
secondary leakage of 1 gpm at room temperature (gpmRT) in a single
SG is used in the Control Rod Ejection (CRE) accident.
No leakage factor will be applied to the locked rotor or control
rod ejection transients due to their short duration.
[[Page 465]]
The TS operational leak rate limit is 150 gallons per day (gpd)
(0.104 gpmRT). The maximum accident leak rate ratio for Salem Unit 1
is 2.16. Consequently, this results in significant margin between
the conservatively estimated accident leakage and the allowable
accident leakage.
For the condition monitoring (CM) assessment, the component of
leakage from the prior cycle from below the H* distance will be
multiplied by a factor of 2.16 and added to the total leakage from
any other source and compared to the allowable accident induced
leakage limit. For the operational assessment (OA), the difference
in the leakage between the allowable leakage and the accident
induced leakage from sources other than the tubesheet expansion
region will be divided by 2.16 and compared to the observed
operational leakage.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change that alters the steam generator inspection
and reporting criteria does not introduce any new equipment, create
new failure modes for existing equipment, or create any new limiting
single failures. Plant operation will not be altered, and all safety
functions will continue to perform as previously assumed in accident
analyses. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed change defines the safety significant portion of
the tube that must be inspected and repaired (plugged). WCAP-17071-P
identifies the specific inspection depth below which any type tube
degradation shown to have no impact on the performance criteria in
[Nuclear Energy Institute (NEI) document] NEI 97-06 [Revision] 2,
``Steam Generator Program Guidelines.''
The proposed change that alters the steam generator inspection
and reporting criteria maintains the required structural margins of
the SG tubes for both normal and accident conditions. Nuclear Energy
Institute 97-06, ``Steam Generator Program Guidelines,'' and NRC
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam
Generator Tubes,'' are used as the bases in the development of the
limited hot leg tubesheet inspection depth methodology for
determining that SG tube integrity considerations are maintained
within acceptable limits. RG 1.121 describes a method acceptable to
the NRC for meeting General Design Criteria (GDC) 14, ``Reactor
Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant System
Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant Pressure
Boundary,'' and GDC 32, ``Inspection of Reactor Coolant Pressure
Boundary,'' by reducing the probability and consequences of a SGTR.
RG 1.121 concludes that by determining the limiting safe conditions
for tube wall degradation, the probability and consequences of a
SGTR are reduced. This RG uses safety factors on loads for tube
burst that are consistent with the requirements of Section III of
the American Society of Mechanical Engineers (ASME) Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, Westinghouse WCAP-17071-P
defines a length of degradation-free expanded tubing that provides
the necessary resistance to tube pullout due to the pressure induced
forces, with applicable safety factors applied. Application of the
limited hot and cold leg tubesheet inspection criteria will preclude
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage as described in
WCAP-1707[1]-P shows that significant margin exists between an
acceptable level of leakage during normal operating conditions that
ensures meeting the accident-induced leakage assumptions and the TS
leakage limit of 150 gpd.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Vincent Zabielski, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Order Imposing Procedures for Access to Sensitive Unclassified Non-
Safeguards Information for Contention Preparation
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant (HBRSEP) Unit No. 2, Darlington County, South Carolina
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station,
Unit No. 1, Salem County, New Jersey
A. This Order contains instructions regarding how potential parties
to this proceeding may request access to documents containing Sensitive
Unclassified Non-Safeguards Information (SUNSI).
B. Within 10 days after publication of this notice of hearing and
opportunity to petition for leave to intervene, any potential party who
believes access to SUNSI is necessary to respond to this notice may
request such access. A ``potential party'' is any person who intends to
participate as a party by demonstrating standing and filing an
admissible contention under 10 CFR 2.309. Requests for access to SUNSI
submitted later than 10 days after publication will not be considered
absent a showing of good cause for the late filing, addressing why the
request could not have been filed earlier.
[[Page 466]]
C. The requestor shall submit a letter requesting permission to
access SUNSI to the Office of the Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, and provide a copy to the Associate General
Counsel for Hearings, Enforcement and Administration, Office of the
General Counsel, Washington, DC 20555-0001. The expedited delivery or
courier mail address for both offices is: U.S. Nuclear Regulatory
Commission, 11555 Rockville Pike, Rockville, Maryland 20852. The e-mail
address for the Office of the Secretary and the Office of the General
Counsel are [email protected] and [email protected],
respectively.\1\ The request must include the following information:
---------------------------------------------------------------------------
\1\ While a request for hearing or petition to intervene in this
proceeding must comply with the filing requirements of the NRC's
``E-Filing Rule,'' the initial request to access SUNSI under these
procedures should be submitted as described in this paragraph.
---------------------------------------------------------------------------
(1) A description of the licensing action with a citation to this
Federal Register notice;
(2) The name and address of the potential party and a description
of the potential party's particularized interest that could be harmed
by the action identified in C.(1);
(3) The identity of the individual or entity requesting access to
SUNSI and the requestor's basis for the need for the information in
order to meaningfully participate in this adjudicatory proceeding. In
particular, the request must explain why publicly available versions of
the information requested would not be sufficient to provide the basis
and specificity for a proffered contention;
D. Based on an evaluation of the information submitted under
paragraph C.(3) the NRC staff will determine within 10 days of receipt
of the request whether:
(1) There is a reasonable basis to believe the petitioner is likely
to establish standing to participate in this NRC proceeding; and
(2) The requestor has established a legitimate need for access to
SUNSI.
E. If the NRC staff determines that the requestor satisfies both
D.(1) and D.(2) above, the NRC staff will notify the requestor in
writing that access to SUNSI has been granted. The written notification
will contain instructions on how the requestor may obtain copies of the
requested documents, and any other conditions that may apply to access
to those documents. These conditions may include, but are not limited
to, the signing of a Non-Disclosure Agreement or Affidavit, or
Protective Order \2\ setting forth terms and conditions to prevent the
unauthorized or inadvertent disclosure of SUNSI by each individual who
will be granted access to SUNSI.
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\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.
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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\
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\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.
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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 23rd day of December 2009.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
Attachment 1--General Target Schedule for Processing and Resolving
Requests for Access to Sensitive Unclassified Non-Safeguards Information
in This Proceeding
------------------------------------------------------------------------
Day Event/activity
------------------------------------------------------------------------
0............................ Publication of Federal Register notice of
hearing and opportunity to petition for
leave to intervene, including order with
instructions for access requests.
10........................... Deadline for submitting requests for
access to Sensitive Unclassified Non-
Safeguards Information (SUNSI) with
information: Supporting the standing of
a potential party identified by name and
address; describing the need for the
information in order for the potential
party to participate meaningfully in an
adjudicatory proceeding.
[[Page 467]]
60........................... Deadline for submitting petition for
intervention containing: (i)
Demonstration of standing; (ii) all
contentions whose formulation does not
require access to SUNSI (+25 Answers to
petition for intervention; +7 requestor/
petitioner reply).
20........................... Nuclear Regulatory Commission (NRC) staff
informs the requestor of the staff's
determination whether the request for
access provides a reasonable basis to
believe standing can be established and
shows need for SUNSI. (NRC staff also
informs any party to the proceeding
whose interest independent of the
proceeding would be harmed by the
release of the information.) If NRC
staff makes the finding of need for
SUNSI and likelihood of standing, NRC
staff begins document processing
(preparation of redactions or review of
redacted documents).
25........................... If NRC staff finds no ``need'' or no
likelihood of standing, the deadline for
requestor/petitioner to file a motion
seeking a ruling to reverse the NRC
staff's denial of access; NRC staff
files copy of access determination with
the presiding officer (or Chief
Administrative Judge or other designated
officer, as appropriate). If NRC staff
finds ``need'' for SUNSI, the deadline
for any party to the proceeding whose
interest independent of the proceeding
would be harmed by the release of the
information to file a motion seeking a
ruling to reverse the NRC staff's grant
of access.
30........................... Deadline for NRC staff reply to motions
to reverse NRC staff determination(s).
40........................... (Receipt +30) If NRC staff finds standing
and need for SUNSI, deadline for NRC
staff to complete information processing
and file motion for Protective Order and
draft Non-Disclosure Affidavit. Deadline
for applicant/licensee to file Non-
Disclosure Agreement for SUNSI.
A............................ If access granted: Issuance of presiding
officer or other designated officer
decision on motion for protective order
for access to sensitive information
(including schedule for providing access
and submission of contentions) or
decision reversing a final adverse
determination by the NRC staff.
A + 3........................ Deadline for filing executed Non-
Disclosure Affidavits. Access provided
to SUNSI consistent with decision
issuing the protective order.
A + 28....................... Deadline for submission of contentions
whose development depends upon access to
SUNSI. However, if more than 25 days
remain between the petitioner's receipt
of (or access to) the information and
the deadline for filing all other
contentions (as established in the
notice of hearing or opportunity for
hearing), the petitioner may file its
SUNSI contentions by that later
deadline.
A + 53....................... (Contention receipt +25) Answers to
contentions whose development depends
upon access to SUNSI.
A + 60....................... (Answer receipt +7) Petitioner/Intervenor
reply to answers.
>A + 60...................... Decision on contention admission.
------------------------------------------------------------------------
[FR Doc. E9-31060 Filed 1-4-10; 8:45 am]
BILLING CODE 7590-01-P