[Federal Register Volume 77, Number 152 (Tuesday, August 7, 2012)]
[Notices]
[Pages 47123-47131]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-19004]
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NUCLEAR REGULATORY COMMISSION
[NRC-2012-0181]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
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) is publishing this regular biweekly 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 or
combined license, as applicable, 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 biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 12, 2012 to July 25, 2012. The last
biweekly notice was published on July 24, 2012 (77 FR 43374).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and are publicly available,
by searching on http://www.regulations.gov under Docket ID NRC-2012-
0181. You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0181. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2012-0181 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and are publicly available, by any of the following methods:
Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0181.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS
[[Page 47124]]
by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2012-0181 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information in comment submissions that you do not want to be publicly
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not edit comment submissions to remove such
information before making the comment submissions available to the
public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined 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.
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.
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 or
combined 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 NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. The NRC regulations are accessible electronically from the NRC
Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to
intervene is filed by the above date, 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 identify 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, 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
[[Page 47125]]
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 the 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 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email 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-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 the 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 the 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's
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 the NRC guidance
available on the NRC's 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 email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's 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's Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email 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, 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 the
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 license amendment
application, see the
[[Page 47126]]
application for amendment which is available for public inspection at
the NRC's PDR, located at One White Flint North, Room O1-F21, 11555
Rockville Pike (first floor), Rockville, Maryland 20852. Publicly
available documents created or received at the NRC are accessible
electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who
encounter problems in accessing the documents located in ADAMS, should
contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737,
or by email to [email protected].
Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South
Carolina
Date of amendment request: June 8, 2012.
Description of amendment request: The proposed change would revise
the Technical Specifications (TSs) 3.1.4, ``Rod Group Alignment
Limits,'' and TS 3.1.7, ``Rod Position Indication,'' to allow up to 1
hour of soak time following substantial rod movement during which
individual rod position indicators may not be within its limits.
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.
This license amendment request proposes to allow up to one hour
of soak time following substantial rod movement during which time
the rod position indication may be outside its limits. This would
allow an additional hour for rod position indication to be
inoperable or a control rod to be misaligned prior to entry into a
TS LCO [Limiting Condition for Operation] Condition and Required
Actions. RPI [Rod Position Indicators] instrumentation is not an
assumed accident initiator; however, the HBRSEP, Unit No. 2 safety
analyses consider two types of rod misalignment events, static
misalignment and a dropped rod.
The safety analyses show that for the static misalignment event,
without any operator intervention, a single fully withdrawn rod
event does not result in any fuel pin failure; therefore, the static
rod misalignment event is not time dependent and an additional hour,
with the misalignment undetected and unmitigated does not increase
the consequences of the event. Multiple rod misalignment events are
bounded by the single rod misalignment analyses and therefore an
additional hour would not have any impact on this event.
The safety analyses also show that a single dropped rod event,
without any operator intervention, does not result in any fuel pin
failure; therefore, the rod drop event is not time dependent and an
additional hour with the misalignment undetected and unmitigated
does not increase the consequences of the event. Multiple rod drop
events cause the reactor to trip and therefore an additional hour
would not have any impact on that event.
Although this license amendment request may allow a misaligned
rod to be undetected for an additional hour, the additional time for
discovery does not change the probability of a misaligned control
rod event because the one hour time extension does not affect the
control rod drive system features that would result in either type
of misalignment.
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 previously evaluated?
Response: No.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. No new accident scenarios, transient
precursors, failure mechanisms, or limiting single failures are
introduced as a result of these changes. No new equipment
performance burdens are imposed.
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
the margin of safety?
Response: No.
The RPI system is an instrumentation system that provides
indication to the operators that a control rod may be misaligned.
Inoperable individual RPI instrumentation does not, by itself in any
way, harm or impact reactor operation. Inoperable rod position
indication may impair the ability of the operators to detect a
misaligned rod. However, the impact of inoperable RPI
instrumentation may be offset by availability of other indications
that a rod is misaligned such as nuclear instrumentation indication
that reactor power has shifted to one side of the core or
thermocouple indication that the core temperatures increased in one
region of the core and/or decreased in another region of the core.
Based on plant experience, the likelihood of a misaligned rod at
HBRSEP, Unit No. 2 is considered to be small and the likelihood of a
misaligned rod coincident with inoperable rod position indication
during the allowed one hour extension is even smaller. In addition,
these proposed changes may enhance plant safety and reliability
because the one hour soak time will allow the operators and
engineers to focus on monitoring the reactor performance without
unnecessary entry into TS LCO Conditions and Required Actions.
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 T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Acting Branch Chief: Jessie F. Quichocho.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: April 30, 2012.
Description of amendment request: This amendment request proposes
to permanently revise technical specification (TS) 6.8.4.j, Steam
Generator (SG) Surveillance Program, to exclude portions of the SG tube
below the top of the SG tubesheet from periodic tube inspections.
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 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 SG tube rupture (SGTR) event
and the steam line break (SLB) postulated accident.
Addressing 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
[[Page 47127]]
tubesheet, and from the differential pressure between the primary
and secondary side, and tubesheet rotation. The structural margins
against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases
for Plugging Degraded PWR [Pressurized-Water Reactors] Steam
Generator Tubes'' [Reference 7] and NEI [Nuclear Energy Institute]
97-06, ``Steam Generator Program Guidelines'', [Reference 3] are
maintained for both normal and postulated accident conditions.
For the portion of the tube outside of the tubesheet, the
proposed change also has no impact on the structural or leakage
integrity. Therefore, the proposed change does not result in a
significant increase in the probability of the occurrence of a SGTR
accident.
At normal operating pressures, leakage from primary water stress
corrosion cracking 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 tube rupture. Therefore, the proposed
change does not result in a significant increase in the consequences
of an SGTR. In addition, the selected H* value envelopes the depth
within the tubesheet required to prevent a tube pullout.
The probability of a SLB is unaffected by the potential failure
of a SG tube as the failure of a tube is not an initiator for a SLB
event.
The leak rate factor of 1.82 for Turkey Point Units 3 and 4, for
a postulated SLB, has been calculated as shown in References 2, 9
and 19. Turkey Point Units 3 and 4 will apply the factor of 1.82 to
the normal operating leakage associated with the tubesheet expansion
region in the condition monitoring (CM) and operational assessment
(OA). 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. Multiplying the TS operational leak rate limit of
150 gpd (at room temperature) through any one SG by a factor of 1.82
shows that the maximum primary to secondary accident induced leak
rate is limited to 273 gpd. This leakage rate is bounded by the
current licensing basis assumed primary to secondary accident leak
rate of 0.20 gpm (288 gpd) through any one SG for SLB. Since the
existing limit on operational leakage continues to ensure that the
SLB assumed accident induced leakage will not be exceeded, the
consequences of a SLB accident are not increased.
For the CM assessment, the component of leakage from the prior
cycle from below the H* distance will be multiplied by a factor of
1.82 and added to the total leakage from any other source and
compared to the allowable accident induced leak rate. For the OA,
the difference in the leakage between the allowable leakage and the
calculated accident induced leakage from sources other than the
tubesheet expansion region will be divided by 1.82 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. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No
The proposed change that alters the SG 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. Does the change involve a significant reduction in a margin
of safety?
Response: No
The proposed change defines the safety significant portion of
the tube that must be inspected and repaired. WCAP-17345, Rev. 2
[Reference 9] identifies the specific inspection depth below which
any type of tube degradation is shown to have no impact on the
performance criteria in NEI 97-06 Rev. 3, ``Steam Generator Program
Guidelines'' [Reference 3] and TS 6.8.4.j, ``Steam Generator (SG)
Program.''
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. Nuclear Energy Institute
97-06, ``Steam Generator Program Guidelines'' [Reference 3], and NRC
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam
Generator Tubes'' [Reference 7], 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 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-17091-P, Rev.
0 [Reference 2] and WCAP-17345, Rev. 2 [Reference 9] define 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 SLB leak rate factor for Turkey Point Units 3 and 4
is 1.82 (Table 9-7 in WCAP-17091-P). Multiplying the TS operational
leak rate limit of 150 gpd through any one SG by the leak rate
factor of 1.82 shows that the maximum primary to secondary accident
induced leak rate is limited to 273 gpd. This leakage rate is
bounded by the current licensing basis assumed primary to secondary
accident leak rate of 0.20 gpm (288 gpd) through any one SG for SLB.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Acting Branch Chief: Jessie F. Quichocho.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 30, 2012.
Description of amendment request: The proposed amendment would
revise Technical Specification Section 2.0, ``Safety Limits.''
Specifically, the proposed amendment would revise two recirculation
loop and single recirculation loop Safety Limit Minimum Critical Power
Ratio (SLMCPR) values to reflect results of a cycle-specific
calculation.
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.
Four accidents have been evaluated previously as reflected in
the CNS [Cooper Nuclear Station] Updated Safety Analysis Report
(USAR). These four accidents are (1) loss-of-coolant, (2) control
rod drop, (3) main steam line break, and (4) fuel handling. The
probability of an evaluated accident is derived from the
probabilities of the
[[Page 47128]]
individual precursors to that accident. Changing the SLMCPR values
does not increase the probability of an evaluated accident. The
change does not require any physical modifications to the plant or
any components, nor does it require a change in plant operation.
Therefore, no individual precursors of an accident are affected.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. This proposed change makes no modification to the
design or operation of the systems that are used in mitigation of
accidents. Limits have been established, consistent with Nuclear
Regulatory Commission (NRC) approved methods, to ensure that fuel
performance during normal, transient, and accident conditions is
acceptable. The proposed change to the values of the SLMCPR
continues to conservatively establish this safety limit such that
the fuel is protected during normal operation and during any plant
transients or anticipated operational occurrences.
Based on the above, NPPD [Nebraska Public Power District]
concludes that the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident from an accident previously evaluated would require
creation of precursors of that accident. New accident precursors may
be created by modification of the plant configuration or changes in
how the plant is operated. The proposed change does not involve a
modification of the plant configuration or in how the plant is
operated. The proposed change to the SLMCPR values assures that
safety criteria are maintained.
Based on the above, NPPD concludes that 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 a
margin of safety?
Response: No.
The values of the proposed SLMCPR provides a margin of safety by
ensuring that no more than 0.1% of fuel rods are expected to be in
boiling transition if the Minimum Critical Power Ratio limit is not
violated. The proposed change will ensure the appropriate level of
fuel protection is maintained. Additionally, operational limits are
established based on the proposed SLMCPR to ensure that the SLMCPR
is not violated during all modes of operation. This will ensure that
the fuel design safety criteria are met (i.e., that at least 99.9%
of the fuel rods do not experience transition boiling during normal
operation as well as anticipated operational occurrences).
Based on the above, NPPD concludes that the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit 1, Washington County, Nebraska
Date of amendment request: February 10, 2012.
Description of amendment request: The proposed amendment would
establish the limiting condition for operation (LCO) requirements for
the reactor protective system (RPS) actuation circuits in Technical
Specification (TS) 2.15, ``Instrumentation and Control Systems.''
Specifically, the proposed change: renumbers LCO 2.15(1) through
2.15(4) to 2.15.1(1) through 2.15.1(4), renumbers LCO 2.15(5) to LCO
2.15.3 with an associated Table 2-6, and implements a new LCO 2.15.2
for the RPS logic and trip initiation channels. The Table of Contents
will also be revised to reflect the renumbering and addition of the LCO
for the RPS logic and trip initiation channels and the new Table 2-6.
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 reactor protective system logic and trip initiation channels
meets Criterion 3 of 10 CFR 50.36 for inclusion into Technical
Specification (TS) as a component that is part of the primary
success path and which functions or actuates to mitigate a design
basis accident or transient. The TSs currently does not have
limiting conditions for operations (LCO) specific for this
circuitry, but does contain surveillance requirements. The addition
of LCOs provides additional restrictions on the operation of the
plant and provides required actions and time limits if these
components are incapable of performing their function. As such, the
proposed change does not increase the probability of an accident.
The proposed changes do not alter the physical design of the RPS, or
any other plant structure, system or component (SSC) at Fort Calhoun
Station (FCS).
The proposed changes conform to the Nuclear Regulatory
Commission's (NRC's) regulatory guidance regarding the content of
plant TS as identified in 10 CFR 50.36 and NRC publication NUREG
1432.
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 TS changes do not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Hence, the proposed changes do not introduce any new
accident initiators, nor do they reduce or adversely affect the
capabilities of any plant structure or system in the performance of
their safety function.
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.
The TS operability requirements for the RPS logic and trip
initiation channels ensure there is adequate components operable to
assure safe reactor operation and are necessary to ensure safety
systems accomplish their safety function for design basis accident
events. The proposed TS would revise the applicability for when the
RPS logic and trip initiation channels are required to be operable
to include whenever control element assemblies (CEAs) are capable of
being withdrawn and the reactor coolant system (RCS) is not at
refueling boron concentration. When the RCS boron concentration is
at refueling boron concentration, or when no more than one trippable
control rod is capable of being withdrawn, the RPS function is
already fulfilled. These proposed TS changes for the RPS are aligned
with the applicability and operability requirements provided in
NUREG 1432.
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 A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power
Station (Zion), Units 1 and 2, Lake County, Illinois
Date of amendment request: May 31, 2012.
[[Page 47129]]
Description of amendment request: The proposed amendments would
approve methods of analysis, use of the upgraded fuel handling building
crane system as a single-failure proof crane, and a NUREG 0612
compliant heavy loads handling program.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The existing DSAR [Defueled Safety Analysis Report] analysis
assumes that a spent fuel cask drop occurs. In this analysis, the
physics of the drop, coupled with concrete bumpers on the cask
loading pit and pool edge were used to demonstrate that a postulated
drop of the spent fuel cask near the Spent Fuel Pool neither
impacted the spent fuel directly nor damaged the pool structure in a
manner that adversely affected the spent fuel, when a cask was to be
handled in the cask loading pit. The proposed License Amendment
Request to operate a single-failure proof Fuel Building Crane
demonstrates that no analysis is required for the cask drop event
based on the design and the associated programmatic controls. A drop
of the spent fuel cask handled with a single-failure proof crane
(designed to ASME NOG-1 [``Rules for Construction of Overhead and
Gantry Cranes (Top Running Bridge, Multiple Girder)''] and compliant
with NUREG-0554 [``Single-Failure-Proof Cranes for Nuclear Power
Plants'', ML110450636]), operated in accordance with the
administrative controls of NUREG-0612 [``Control of Heavy Loads at
Nuclear Power Plants,'' ML070250180] has an acceptably low
probability so as to effectively preclude consideration of the
event. The risk of such a drop event using the new single-failure
proof crane operated in accordance with the Heavy Loads Program
procedures, qualitatively, is lower than the event previously
analyzed which postulate the event without evaluation of its
likelihood.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The location and design functions of the Fuel Building crane are
not changed from those currently described in the DSAR. Because the
new crane has a single-failure proof design the uncontrolled
lowering, or drop, of a heavy load will not be considered credible.
Evaluations show that individual malfunctions or component failures
of the crane will not result in load drop. The new single-failure
proof crane['s] primary use[s] will be to move a loaded or unloaded
MAGNASTOR transfer cask between the cask loading pit [and] the
decontamination pit, and transfer [the cask] to the low profile cart
rail transport in the Fuel Handling Building. No components that are
classified as Important to the Defueled Condition, other than the
Fuel Building crane, will be affected by these movements. Based on
the design and programmatic controls on the crane, no load will
lower uncontrollably or drop in or around the spent fuel pool or
near an open cask containing spent fuel nor will a cask containing
spent fuel drop or be lowered uncontrollably during operation of the
crane. Hence no new accidents will be initiated.
Therefore, the proposed change will 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.
This proposed License Amendment Request involves the replacement
of the existing non-single-failure proof Fuel Building Crane with a
new single-failure proof crane. The new crane has been designed to
meet the specifications found in ASME NOG-1-2004, which has been
endorsed by the NRC in RIS 2005-25, as supplemented, as an
acceptable means of meeting the criteria in NUREG-0554, ``Single-
failure Proof Cranes for Nuclear Power Plants.'' to provide adequate
protection and safety margin against the uncontrolled lowering of
the lifted load. The occurrence of a cask load drop accident is
considered not credible when the load is lifted with a single-
failure proof lifting system meeting the guidance in NUREG-0612,
``Control of Heavy Loads at Nuclear Power Plants'' Section 5.1.6,
``Single-Failure-Proof Handling Systems.'' As a result, the proposed
change, replacing the existing non-single-failure proof crane, has
no adverse impact on stored spent fuel, or structural integrity of
the pool.
The configuration of the crane and the primary load, a spent
fuel cask containing spent fuel, is changed from that of the DSAR.
The specific analysis dealing with a drop of the cask will no longer
be applicable and [will be] removed from the DSAR, since the new
single-proof crane makes that event of low enough probability to not
be considered credible. The maximum critical lift capacity of the
crane has not been changed, though the load to be lifted is larger.
The structural analyses of the crane and its support structure,
however, show acceptable margin under the acceptance criteria of
NOG-I for operation of the crane.
Therefore, the proposed change does 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: Russ Workman, Deputy General Counsel,
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT
84101.
NRC Branch Chief: Bruce Watson.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available documents created or received at the
NRC are accessible electronically through the Agencywide Documents
Access and Management System (ADAMS) in the NRC Library at 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's Reference staff at 1-800-397-4209, 301-415-4737 or by
email to [email protected].
[[Page 47130]]
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: July 21, 2011.
Brief description of amendments: The amendments revised Technical
Specifications 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' 3.5.4, ``Refueling Water Storage Tank
(RWST),'' and 3.6.6, ``Containment Spray System.''
Date of issuance: July 25, 2012.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-269 and Unit 2-265.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 20, 2012 (77 FR
16274).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 25, 2012.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Units 2 and 3 (IP2 and IP3), Westchester
County, New York
Date of application for amendment: July 8, 2009, as supplemented by
letters dated September 28, 2009, October 26, 2009, October 5, 2010,
October 28, 2010, July 28, 2011, August 23, 2011, October 28, 2011,
December 15, 2011, January 11, 2012, March 2, 2012, April 23, 2012, and
May 7, 2012.
Brief description of amendment: The amendment authorizes the
transfer of spent fuel from the IP3 spent fuel pool to the IP2 spent
fuel pool, using a newly-designed shielded transfer canister, for
further transfer to the on-site Independent Spent Fuel Storage
Installation.
Date of issuance: July 13, 2012.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 268 and 246.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the License and the Technical Specifications.
Date of initial notice in Federal Register: January 21, 2010 (75 FR
3497).
The supplements provided additional information that clarified the
application but did not expand the scope of the application as
originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 13, 2012.
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 application for amendment: September 8, 2010, as
supplemented by letters dated November 18, 2010, November 23, 2010,
February 23, 2011 (four letters), March 9, 2011 (two letters), March
22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April 21,
2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June 15,
2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011, August
25, 2011, August 29, 2011, August 30, 2011, September 2, 2011,
September 9, 2011, September 12, 2011, September 15, 2011, September
26, 2011, October 10, 2011, October 24, 2011, November 14, 2011,
November 25, 2011, November 28, 2011, December 19, 2011, February 6,
2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21,
2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May
9, 2012, and June 12, 2012.
Brief description of amendment: The amendment increased the maximum
steady-state reactor core power level from 3,898 megawatts thermal
(MWt) to 4,408 MWt, which is an increase of approximately 15 percent
from the original licensed thermal power level of 3,833 MWt. The
proposed increase in power level is considered an extended power
uprate.
Date of issuance: July 18, 2012.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No: 191.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 10, 2011 (76 FR
1464). The supplemental letters dated November 18, 2010, November 23,
2010, February 23, 2011 (four letters), March 9, 2011 (two letters),
March 22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April
21, 2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June
15, 2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011,
August 25, 2011, August 29, 2011, August 30, 2011, September 2, 2011,
September 9, 2011, September 12, 2011, September 15, 2011, September
26, 2011, October 10, 2011, October 24, 2011, November 14, 2011,
November 25, 2011, November 28, 2011, December 19, 2011, February 6,
2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21,
2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May
9, 2012, and June 12, 2012, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 18, 2012.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410,
Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York
Date of application for amendments: July 20, 2011, as supplemented
on November 3, 2011, and January 12, 2012.
Brief description of amendments: The amendments revised the NMP1
Technical Specification (TS) Section 5.1, ``Site,'' and associated TS
Figure 5.1-1, ``Site Boundaries, Nine Mile Point-Unit 1,'' and the NMP2
TS Figure 4.1-1, ``Site Area and Land Portion of Exclusion Area
Boundaries,'' to reflect the transfer of a portion of the Nine Mile
Point Nuclear Station, LLC (NMPNS) site real property located outside
of the NMPNS Protected Area but within the current NMPNS Owner
Controlled Area, as well as specified easements over the remainder of
the NMPNS site, to Nine Mile Point 3 Nuclear Project, LLC (NMP3), a
subsidiary of UniStar Nuclear Energy, LLC.
Date of issuance: July 12, 2012.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 212 for Unit 1 and 142 for Unit 2.
Renewed Facility Operating License Nos. DPR-63 and NPF-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: December 27, 2011 (76
FR 80977).
The supplements dated November 3, 2011, and January 12, 2012,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the Nuclear Regulatory Commission (NRC) staff's initial proposed
no significant hazards consideration determination noticed in
[[Page 47131]]
the Federal Register on December 27, 2011 (76 FR 80977).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 12, 2012.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 26th day of July 2012.
For the Nuclear Regulatory Commission.
Louise Lund,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2012-19004 Filed 8-6-12; 8:45 am]
BILLING CODE 7590-01-P