[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]
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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
[email protected], 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
[email protected], 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 [email protected].
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 [email protected] and [email protected],
respectively.\1\ The request must include the following information:
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\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\
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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 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.
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[FR Doc. 2011-2027 Filed 1-26-11; 4:15 pm]
BILLING CODE 7590-01-P