[Federal Register Volume 75, Number 133 (Tuesday, July 13, 2010)]
[Notices]
[Pages 39975-39985]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-16879]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0248]
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations; Biweekly Notice
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) 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
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 June 17, 2010, to June 30, 2010. The last
biweekly notice was published on June 29, 2010 (75 FR 37471).
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), Sec. 50.92, this means that operation of the
facility in accordance with the proposed amendment would not: (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, 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, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
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/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
[[Page 39976]]
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 held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, any hearing held
would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the participant should
contact the Office of the Secretary by e-mail at
[email protected], or by telephone at (301) 415-1677, to request:
(1) A digital ID certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through EIE, users will be required to install a Web
browser plug-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. e.t. on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., e.t.,
Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
social security numbers, home addresses, or home phone numbers in
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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 application for amendment which is available for
public inspection at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, 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 PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].
Duke Energy Carolinas, LLC, et al., Docket No. 50-414, Catawba Nuclear
Station, Unit 2, York County, South Carolina
Date of amendment request: April 28, 2010.
Description of amendment request: The amendment would revise
Technical Specification (TS) 5.5.9 to exclude portions of the Steam
Generator (SG) tube from periodic SG tube inspections and plugging or
repair. In addition, reporting requirement changes are proposed to TS
5.6.8. This submittal is requesting a one-cycle approval for the
Catawba Nuclear Station, Unit 2, End of Cycle 17 Refueling Outage and
subsequent Cycle 18 operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 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 TS 5.5.9, TS 5.6.8, and
the Facility Operating License have no significant effect upon
accident probabilities or consequences. Of the various accidents
previously evaluated, the following are limiting with respect to the
proposed changes as discussed in this amendment request:
SG Tube Rupture evaluation
Steam Line Break/Feed Line Break evaluation
Locked Rotor evaluation
Control Rod Ejection evaluation
Loss of Coolant Accident conditions cause a compressive axial
load to act on the tube. Therefore, since this accident 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; however, the seismic
analysis of Model D5 SGs (the SGs at Catawba) has shown that axial
loading of the tubes is negligible during this event.
At normal operating pressures, leakage from Primary Water Stress
Corrosion Cracking (PWSCC) below 16.95 inches from the top of the
tubesheet is limited by both the tube-to-tubesheet crevice and the
limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region.
For the SG Tube Rupture event, tube rupture is precluded for
cracks in the hydraulic expansion region due to the constraint
provided by the tubesheet. Therefore, the margin against tube burst/
pullout is maintained during normal and postulated accident
conditions and the proposed change does not result in a significant
increase in the probability of a tube rupture. SG Tube Rupture
consequences 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 that from a severed tube. Therefore, the
proposed change does not result in a significant increase in the
consequences of a tube rupture.
The probability of a Steam Line Break/Feed Line Break, Locked
Rotor, and Control Rod Ejection are not affected by the potential
failure of a SG tube, as the failure of a tube is not an initiator
for any of these events. In WCAP-17072-P, leakage is modeled as flow
through a porous medium via 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. The fluid
viscosity is based on fluid temperature and it has been shown that
for the most limiting accident, the fluid temperature does not
exceed the normal operating temperature. Therefore, the viscosity
ratio is assumed to be 1.0 and 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 2.65 for Catawba Unit 2 for a postulated
Steam Line Break/Feed Line Break has been calculated as shown in
WCAP-17072-P, as supplemented. The leakage factor has been increased
to 3.27 per additional Westinghouse analysis specific to Catawba.
Therefore, Catawba Unit 2 will apply a factor of 3.27 to the normal
operating leakage associated with the tubesheet expansion region in
the Condition Monitoring assessment and Operational Assessment.
Through application of the limited tubesheet inspection scope, the
proposed operating leakage limit provides assurance that excessive
leakage (i.e., greater than accident analysis assumptions) will not
occur. No leakage factor will be applied to the Locked Rotor or
Control Rod Ejection due to their short duration, since the
calculated leak rate ratio is less than 1.0. Therefore, the proposed
change does not result in a significant increase in the consequences
of these accidents.
For the Condition Monitoring assessment, the component of
leakage from the prior cycle from below the H* distance will be
multiplied by a factor of 3.27 and added to the total leakage from
any other source and compared to the allowable accident induced
leakage limit. For the Operational Assessment, 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 3.27 and compared to the observed operational leakage.
Based on the above, the performance criteria of NEI 97-06,
Revision 2 and RG [Regulatory Guide] 1.121 continue to be met and
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 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 TS 5.5.9, TS 5.6.8, and
the Facility Operating License do 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 one-cycle 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 change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No. The proposed changes to TS 5.5.9, TS 5.6.8, and
the Facility Operating License maintain the required structural
margins of the SG tubes for both normal and accident conditions. NEI
97-06, Revision 2 and RG 1.121 are used as the basis 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
[[Page 39978]]
method acceptable to the NRC staff for meeting GDC [General Design
Criteria] 14, 15, 31, and 32 by reducing the probability and
consequences of a SG Tube Rupture. RG 1.121 concludes that by
determining the limiting safe conditions for tube wall degradation,
the probability and consequences of a SG Tube Rupture are reduced.
This RG uses safety factors on loads for tube burst that are
consistent with the requirements of Section III of the ASME
[American Society of Mechanical Engineers] Code [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, WCAP-17072-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 shows that significant margin exists between an
acceptable level of leakage during normal operating conditions that
ensures meeting the accident induced leakage assumption and the TS
leakage limit.
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: Ms. Kate Nolan, Associate General Counsel
and Managing Attorney, Duke Energy Carolinas, LLC, 422 South Church
Street, Mail Code--EC07H, P.O. Box 1244, Charlotte, NC 28201-1244.
NRC Branch Chief: Gloria Kulesa.
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: June 25, 2009, as supplemented May 21,
2010.
Description of amendment request: Revise the licensing bases to
adopt the alternative source term as allowed in Title 10 of the Code of
Federal Regulations, Sec. 50.67.
An application that addressed similar issues was previously
submitted on June 25, 2009, and noticed in the Federal Register (FR) on
December 29, 2009 (74 FR 68870). Due to certain changes in the
specifics stated in the May 21, 2010, supplement, from those proposed
in the June 25, 2009, application, this is a renotice that includes
those changes. Below is the no significant hazards consideration
determination (NSHCD) for the changes in the May 21, 2010 supplemental.
The original NSHCD as published in the FR December 29, 2009, still
applies to the June 25, 2009 application.
Basis for proposed NSHCD: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes to [technical specification] TS 4.4.8 will
only provide for better assurance of required sampling and analysis
of the reactor coolant system specific activity during thermal power
changes and transient conditions (MODES 1, 2, 3, and 4). This will
ensure potential consequences of a [defined-basis accident] DBA are
bounded by the approved accident analyses.
The proposed changes to TS 3/4.7.5 itemize the system
operability requirements and appropriate actions in the event that
those requirements are not satisfied. These actions include actions
to be taken during the allowed outage times (AOTs) specified in the
actions to bring the system back into compliance with the system
operability requirements. The actions also provide for restoration
of the inoperable component or in some cases provide for placing and
maintaining it in a safe condition until it can be restored. The
actions may include compensatory measures that require initiation of
mitigating actions involving operator action to manually align and
place into service a compensatory filtration unit in the event that
the normal filtration train is out-of-service. These compensatory
measures are required to be taken within 24 hours compared to the
current allowed outage time of 84 hours for system inoperability
without any compensatory measures specified. Moreover, consistent
with the current Turkey Point TS and TSTF-448 AOTs, manually
aligning the compensatory filter within 24 hours to maintain
[control room emergency ventilation system] CREVS operability is
acceptable in order to ensure control room operations will be
protected from analyzed radiological hazards. The other action
statements for inoperability of a redundant active component provide
for an AOT of 7 days consistent with the Westinghouse Standard
Technical Specification. They are based on the low probability of
occurrence of a DBA challenging the Control Room Habitability during
this time period and the continued capability of the remaining
system components to perform the required CREVS safety function.
The proposed changes have no effect on the probability of an
accident previously evaluated as they do not affect any accident
initiators. The proposed changes have no significant effect on the
consequences of an accident previously evaluated as they either
provide for better monitoring of plant operating parameters or for
compensatory actions to be taken for out-of-service equipment not
previously available. Design changes to enhance the system
capabilities will be made to the same design and quality standards
as the existing CREVS. System modifications required to support
these proposed changes are evaluated under the 10 CFR 50.59 program
and are enhancements to the mitigation strategies.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes to TS 4.4.8 will only provide for better
assurance of required sampling and analysis of the reactor coolant
system specific activity during MODES 1, 2, 3, and 4. The proposed
modifications to the plant configuration will be fully qualified to
the appropriate design requirements to assure their required
function is available for accident mitigation. Additionally,
functions of other equipment required for accident mitigation are
also not adversely impacted. Design changes to enhance the system
capabilities will be made to the same design and quality standards
as the existing CREVS. The proposed changes to TS 3/4.7.5 will
provide for better specification of system operability requirements
and appropriate actions in the event that those requirements are not
satisfied. The proposed changes have no effect on accident
precursors or initiators and only enhance mitigation capabilities
with regard to protecting control room personnel from radiological
hazards.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes to TS 4.4.8 will only provide for better
assurance of required sampling and analysis of the reactor coolant
system specific activity during thermal power changes and transient
conditions (MODES 1, 2, 3, and 4). No plant system or component
design or operational requirements are affected by these changes.
The proposed changes to TS 3/4.7.5 will provide for better
specification of system operability requirements and appropriate
actions in the event that those requirements are not satisfied. The
proposed increase in the specified AOT for inoperability of CREVS
components from 84 hours to 7 days is considered insignificant as it
is consistent with the Westinghouse Standard Technical Specification
and based on the low probability of occurrence of a DBA challenging
the Control Room Habitability during this time period and the
continued capability of the remaining system components to perform
the required CREVS safety function. Moreover, consistent with the
current Turkey Point TS and TSTF-448 AOTs, manually aligning the
compensatory filter within 24 hours to maintain CREVS operability is
an acceptable margin of safety to ensure control room operations
will be protected from analyzed radiological hazards. The proposed
changes provide for compensatory actions to be taken for out-of-
service equipment that were not previously
[[Page 39979]]
available and thus enhance existing mitigation capabilities with
regard to protecting control room personnel from radiological
hazards.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based on the above discussion, FPL has determined that the
proposed change does not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of Sec. 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 Branch Chief: Douglas A. Broaddus.
NextEra Energy Seabrook, LLC Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham County, New Hampshire
Date of amendment request: May 14, 2010.
Description of amendment request: The proposed changes would revise
the Seabrook Station Technical Specifications (TSs) governing the
Containment Enclosure Emergency Air Cleanup System. Specifically, the
proposed change would insert a requirement that if both trains of the
system are inoperable, at least one train must be returned to operable
status within 24 hours or begin a shutdown of the reactor. Currently,
since there are no limiting conditions for operation proscribed actions
in the event two trains are inoperable, TS 3.0.3 requires a shutdown
within 6 hours.
Basis for proposed NSHC determination: As required by 10 CFR
50.91(a), the licensee has provided its analysis of the issue of NSHC,
which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not impact the physical function of
plant structures, systems, or components (SSCs) or the manner in
which SSCs perform their design function. The proposed changes
neither adversely affect accident initiators or precursors, nor
alter design assumptions. The proposed changes do not alter or
prevent the ability of operable [SSCs] to perform their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits.
This change is a revision to the technical specifications
(TS[s]) for the containment enclosure emergency air cleanup system
(CEEACS), which is a mitigation system designed to prevent
uncontrolled releases of radioactivity into the environment. The
change would allow intermittent opening of the containment enclosure
boundary under administrative controls. These controls would ensure
that the opening will be quickly sealed to maintain the validity of
the licensing basis analyses of accident consequences. The proposed
change adds a new action requirement that would allow 24 hours to
restore the containment enclosure boundary in the event that both
trains of the CEEACS are inoperable due to an inoperable containment
enclosure boundary. The proposed 24 hour completion time is
reasonable based on the low probability of a design basis accident
occurring during this time period and the use of preplanned
compensatory measures. The CEEACS is not an initiator or precursor
to any accident previously evaluated. Therefore, the probability of
any accident previously evaluated is not increased.
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 will not impact the accident analysis. The
changes will not alter the requirements of the CEEACS or its
function during accident conditions, and no new or different
accidents result from the proposed changes to the TS[s]. The changes
do not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the method of plant operation. The changes do not alter
assumptions made in the safety analysis. Therefore, this request
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed changes do not
involve a significant change in the method of plant operation, and
no accident analyses will be affected by the proposed changes.
Additionally, the proposed changes will not relax any criteria used
to establish safety limits, will not relax any safety system
settings, and will not relax the bases for any limiting conditions
for operation. The safety analysis acceptance criteria are not
affected by this change. The proposed change will not result in
plant operation in a configuration outside the design bases. The
proposed change does not adversely affect systems that respond to
safely shutdown the plant and to maintain the plant in a safe
shutdown condition. Therefore, these 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, and with the changes noted above, it appears that the
three standards of Sec. 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves NSHC.
Attorney for licensee: M.S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Harold K. Chernoff.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: March 18, 2010.
Description of amendment request: The proposed amendment would
revise the NMP1 Technical Specifications (TSs) for inoperable snubbers
by removing TS 3/4.6.4, ``Shock Suppressors (Snubbers),'' and would
also add a new Limiting Condition for Operation (LCO) 3.0.8.
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 change to relocate TS 3/4.6.4 to station procedures
is administrative in nature and does not involve the modification of
any plant equipment or affect basic plant operation. Snubber
operability and surveillance requirements will be contained in the
station procedures to ensure design assumptions for accident
mitigation are maintained.
The proposed change to add LCO 3.0.8 allows a delay time for
entering a supported system TS when the inoperability is due solely
to an inoperable snubber if risk is assessed and managed. Entrance
into TS actions or delaying entrance into actions is not an
initiator of any accident previously evaluated. Consequently, the
probability of an accident previously evaluated is not significantly
increased. The consequences of an accident while relying on
allowance provided by proposed LCO 3.0.8 are no different than the
consequences of an accident while relying on the current TS required
actions in effect without the allowance provided by proposed LCO
3.0.8.
Revision of TS Table of Contents to reflect deletion of TS 3/
4.6.4 is administrative in nature and therefore does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
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
[[Page 39980]]
accident from any accident previously evaluated?
Response: No.
The proposed change to relocate TS 3/4.6.4 to station procedures
is administrative and does not involve any physical alteration of
plant equipment. The proposed change does not change the method by
which any safety related system performs its function. As such, no
new or different types of equipment will be installed, and the basic
operation of installed equipment is unchanged. The methods governing
plant operation and testing remain consistent with current safety
analysis assumptions. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed change to add LCO 3.0.8 does not involve a physical
alteration of the plant (no new or different type of equipment will
be installed). Allowing delay times for entering supported system
TSs when inoperability is due solely to inoperable snubbers, if risk
is assessed and managed, will not introduce new failure modes or
effects.
Revision of TS Table of Contents to reflect deletion of TS 3/
4.6.4 is administrative in nature and therefore does 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.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to relocate TS 3/4.6.4 to station procedures
is administrative in nature, does not negate any existing
requirement, and does not adversely affect existing plant safety
margins or the reliability of the equipment assumed to operate in
the safety analysis. As such, there are no changes being made to
safety analysis assumptions, safety limits or safety system settings
that would adversely affect plant safety as a result of the proposed
change. Margins of safety are unaffected by requirements that are
retained, but relocated from the TSs to station procedures.
The proposed change to add LCO 3.0.8 to TSs allows a delay time
before declaring supported TS systems inoperable when the associated
snubber(s) cannot perform the required safety function. The proposed
change retains an allowance in the current NMPI TSs while upgrading
it to be more conservative for snubbers supporting multiple trains
or sub-systems of an associated system. The updated TS will continue
to provide an adequate margin of safety for plant operation upon
incorporation of LCO 3.0.8. The station design and safety analysis
assumptions provide margin in the form of redundancy to account for
periods of time when system capability is reduced.
Revision of TS Table of Contents to reflect deletion of TS 3/
4.6.4 is administrative in nature and therefore does not involve a
significant reduction in a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Carey W. Fleming, Senior Counsel,
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite
200C, Baltimore, MD 21202.
NRC Branch Chief: Nancy L. Salgado.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: March 22, 2010.
Description of amendment request: The proposed amendment would
revise the NMP1 Technical Specifications (TSs) section 4.3.7
``Containment Spray System,'' by modifying the testing frequency for
the Surveillance Requirement (SR) 4.3.7.b, ``Nozzles,'' from ``at least
once per operating cycle * * * '' to ``following maintenance that could
result in nozzle blockage.'' Additional wording changes would be made
to the SR to make it more consistent with the corresponding Standard
TS, SR 3.6.1.7.4.
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 change modifies the SR to verify that the
Containment Spray System (CSS) drywell and torus spray nozzles are
unobstructed after maintenance that could introduce material
resulting in nozzle blockage. The requirement to test the headers
will be removed as well as the type of test to be used. Since the
opening within the pipes is much larger than the nozzles, they are
not likely to become obstructed unless the nozzles become
obstructed. The spray nozzles and headers are not assumed to be
initiators of any previously analyzed accident. Therefore, the
proposed change does not increase the probability of any accident
previously evaluated. The spray nozzles are used in the accident
analyses to mitigate design basis accidents. The revised SR to
verify system operability following maintenance is considered
adequate to ensure operability of the CSS. Since the system will
still be able to perform its accident mitigation function, the
consequences of accidents previously evaluated are not increased.
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 revises the SR to verify that the CSS
nozzles are unobstructed after maintenance that could result in
nozzle blockage. The requirement to test the headers will be removed
as well as the type of test to be used. The spray nozzles and
headers are not assumed to be initiators of any previously analyzed
accident. The change does not introduce a new mode of plant
operation and does not involve a physical modification to the plant.
The change will not introduce new accident initiators or impact the
assumptions made in the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises the frequency for performance of the
SR to verify that the CSS nozzles are unobstructed. The frequency is
changed from ``once per operating cycle'' to ``following maintenance
that could result in nozzle blockage.'' The requirement to test the
headers will be removed as well as the type of test to be used. The
revised testing requirement, along with the foreign material
exclusion program, the normal environmental conditions for the
system, and the remote physical location of the spray nozzles,
provide assurance that the spray nozzles and headers will remain
unobstructed. As the spray nozzles and headers are expected to
remain unobstructed and able to perform their post-accident
mitigation function, plant safety is not significantly affected.
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: Carey W. Fleming, Senior Counsel,
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite
200C, Baltimore, MD 21202.
NRC Branch Chief: Nancy L. Salgado.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: March 30, 2010, as supplemented on June
1, 2010.
[[Page 39981]]
Description of amendment request: The proposed amendment would
revise the NMP2 Technical Specification (TS) section 3.8.1, ``AC
Sources--Operating,'' to extend the Completion Time (CT) for an
inoperable Division 1 or Division 2 diesel generator (DG) from 72 hours
to 14 days.
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 TS change to increase the CT for an inoperable
Division 1 or Division 2 DG from 72 hours to 14 days does not affect
the design, function, operational characteristics, or reliability of
the DGs. The DGs are designed to mitigate the consequences of
previously evaluated accidents and, as such, are not accident
initiators.
Extending the CT for an inoperable DG will not significantly
affect the capability of the DGs to perform their accident
mitigation safety functions or adversely affect DG or offsite power
availability. The consequences of previously evaluated accidents
will not be significantly affected since the remaining DGs
supporting the redundant Engineered Safety Feature (ESF) systems
will continue to be available to perform the accident mitigation
functions as designed.
Both a deterministic evaluation and a risk impact assessment
were performed to support the proposed DG CT extension. The
deterministic evaluation concluded that the defense-in-depth
philosophy will be maintained with the proposed DG CT extension. The
current TS and 10 CFR 50.65 (Maintenance Rule) programmatic
requirements and additional administrative controls provide
assurance that a loss of offsite power occurring concurrent with an
inoperable DG will not result in a complete loss of function of
critical systems. The duration of the proposed DG CT is determined
considering that there is a minimal possibility that an accident
will occur while a component is removed from service. A risk impact
assessment was performed which concluded that the increase in plant
risk due to the increased DG CT is small and consistent with the
guidance contained in Regulatory Guide 1.177, ``An Approach for
Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications.''
Based on the above discussion, it is concluded that the proposed
amendment 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 amendment does not alter the design, configuration,
or method of operation of the plant, and does not alter any safety
analysis inputs or assumptions. The proposed extended DG CT will not
reduce the number of DGs below the minimum required for safe
shutdown or accident mitigation. No new component failure modes,
system interactions, or accident responses will be created that
could result in a new or different kind of accident from any
accident previously evaluated.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed extension of the DG CT remains consistent with
codes and standards applicable to the onsite alternating current
(AC) sources, except that the extension deviates from the
recommendations of Regulatory Guide 1.93, ``Availability of Electric
Power Sources.'' The proposed amendment is justified based on the
results of a deterministic evaluation and a risk impact assessment.
These demonstrate that the defense-in-depth philosophy will be
maintained and the increase in plant risk is small and consistent
with the guidance contained in Regulatory Guide 1.177.
The DG reliability and availability are monitored and evaluated
with respect to Maintenance Rule performance criteria to assure DG
out of service times do not degrade operational safety over time.
Furthermore, extension of the DG CT does not affect any safety
analysis inputs or assumptions and will not erode the reduction in
severe accident risk that was achieved with implementation of the
Station Blackout (SBO) rule (10 CFR 50.63). The SBO coping analysis
is unaffected by the CT extension since the DGs are not assumed to
be available during the coping period. The assumptions used in the
coping analysis regarding DG reliability are unaffected since
preventive maintenance and testing will continue to be performed to
maintain the reliability assumptions.
Based on the above discussion, it is concluded that the proposed
amendment 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: Carey W. Fleming, Senior Counsel,
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite
200C, Baltimore, MD 21202
NRC Branch Chief: Nancy L. Salgado.
Notice of Issuance of Amendments to Facility Operating 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.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, 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, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management System (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: May 29, 2009, as supplemented by
letters dated November 14, 2008, December 11, 2008, August 13, 2009,
August 28, 2009, October 9, 2009, February 4, 2010, and April 5, 2010.
[[Page 39982]]
Brief description of amendment: The proposed amendment transitions
the existing fire protection program to a risk-informed, performance-
based program based on National Fire Protection Association Standard
805 (NFPA 805), ``Performance-Based Standard for Fire Protection for
Light Water Reactor Electric Generating Plants,'' 2001 Edition, in
accordance with Title 10 of the Code of Federal Regulations, Sec.
50.48(c). NFPA 805 allows the use of performance-based methods, such as
fire modeling and fire risk evaluations, to demonstrate compliance with
the nuclear safety performance criteria.
Date of issuance: June 28, 2010.
Effective date: Effective as of the date of issuance and shall be
implemented within 180 days, contingent upon completion of the items
identified in section 2.9 of the associated NRC Safety Evaluation.
Amendment No.: 133.
Renewed Facility Operating License No. NPF-63: The amendment
revises the Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: June 19, 2009 (74 FR
29241). The supplements dated November 14, 2008, December 11, 2008,
August 13, 2009, August 28, 2009, October 9, 2009, February 4, 2010,
and April 5, 2010 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.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment and final NSHC
determination are contained in a safety evaluation dated June 28, 2010.
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: October 2, 2008, as
supplemented by letters dated August 25, 2009, and October 23, 2009.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) associated with the verification of ice
condenser door operability and TS surveillance requirements 3.6.13.5
and 3.6.13.6.
Date of issuance: June 28, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 256 and 251.
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 8, 2010 (75 FR
10513). The supplements dated August 25, 2009, and October 23, 2009,
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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 28, 2010.
No significant hazards consideration comments received: No.
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: September 2, 2008, as
supplemented by letters dated June 18, 2009, July 8, 2009, August 13,
2009, September 8, 2009, November 10, 2009 and March 8, 2010.
Brief description of amendments: The amendments revised the
technical specifications to allow manual operation of the containment
spray system and to revise the upper and lower limits of the refueling
water storage tank.
Date of issuance: June 28, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 257 and 252.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the licenses and the Technical Specifications.
Date of initial notice in Federal Register: April 7, 2009 (74 FR
15767). The supplements dated June 18, 2009, July 8, 2009, August 13,
2009, September 8, 2009, November 10, 2009, and March 8, 2010, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 28, 2010.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: October 2, 2008, as
supplemented by letters dated August 25, 2009, and October 23, 2009.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) associated with the verification of ice
condenser door operability and TS surveillance requirements 3.6.13.5
and 3.6.13.6.
Date of issuance: June 28, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 256 and 236.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 8, 2010 (75 FR
10508). The supplements dated August 25, 2009, and October 23, 2009,
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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 28, 2010.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: July 23, 2009.
Brief description of amendment: The amendment removed the local
refueling water storage tank level indication from Technical
Specification Surveillance Requirement 3.5.4.5.
Date of issuance: June 28, 2010.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 241.
Facility Operating License No. DPR-64: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: October 6, 2009 (74 FR
51329).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 2010.
No significant hazards consideration comments received: No.
[[Page 39983]]
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: June 30, 2009, as supplemented
by letter dated May 24, 2010.
Brief description of amendment: This amendment revises the
Surveillance Requirement (SR) regarding the start time tests for the
Division 3 Emergency Diesel Generator to provide consistency with
existing similar Technical Specification (TS) 3.8.1 ``AC Sources--
Operating'' SRs and the time provided in the licensing basis emergency
core cooling system analyses.
Date of issuance: June 30, 2010.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 154.
Facility Operating License No. NPF-58: This amendment revised the
TSs and License.
Date of initial notice in Federal Register: November 17, 2009 (74
FR 59261). The supplement dated May 24, 2010 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 June 30, 2010.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Units 1 and 2 (CNP-1 and CNP-2), Berrien
County, Michigan
Date of application for amendment: January 14, 2009 as supplemented
by letters dated October 30, 2009, and March 19, 2010.
Brief description of amendment: The amendment modifies the
Operating License, Condition 2.C.(2), Appendix B, Environmental
Technical Specifications, Part II, ``Non-Radiological Environmental
Protection Plan.'' The amendment deletes outdated program information
and relieves I&M from preparing and submitting unnecessary or
duplicative environmental reports.
Date of issuance: June 24, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 312 (CNP-1), 295 (CNP-2).
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Renewed Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 5, 2009 (74 FR
20749).
The supplemental information dated October 30, 2009, and March 19,
2010, contained clarifying information, did not change the scope of
January 14, 2009, application or the initial no significant hazards
consideration determination, and does not expand the scope of the
original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 24, 2010.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2 (NMP2), Oswego County, New York
Date of application for amendment: December 18, 2009.
Brief description of amendment: The amendment changes the NMP2
Technical Specifications (TSs) for unavailable barriers by adding
Limiting Condition for Operation (LCO) 3.0.9. LCO 3.0.9 establishes
conditions under which a supported system would remain operable when
required physical barriers are not capable of providing their related
support function. The submitted change is consistent with the industry
Technical Specifications Task Force (TSTF) Traveler TSTF-427, Revision
2, ``Allowance for Non Technical Specification Barrier Degradation on
Supported System OPERABILITY.'' A notice of the TSTF-427, Revision 2 TS
improvement was published in the Federal Register on October 3, 2006
(71 FR 58444) as part of the Consolidated Line Item Improvement
Process.
Date of issuance: June 29, 2010.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 135.
Renewed Facility Operating License No. NPF-069: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: April 6, 2010 (75 FR
17445).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 2010.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1 (NMP1), Oswego County, New York
Date of application for amendment: July 2, 2009.
Brief description of amendment: The amendment revises the TSs by
removing position indication for the relief valves and safety valves
from TS 3.6.11, ``Accident Monitoring Instrumentation.'' The amendment
would also correct an editorial error in the title of Table 4.6.11,
``Accident Monitoring Instrumentation Surveillance Requirement.''
Date of issuance: June 29, 2010.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 205.
Renewed Facility Operating License No. NPF-069: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: October 14, 2009 (74 FR
52826).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 2010.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of application for amendments: November 4, 2008, as
supplemented by letters dated August 10, 2009, and March 30, 2010.
Brief description of amendments: The amendments modify the
technical specifications (TSs) and facility operating licenses by
increasing the 24-month test load for the Unit 1 emergency diesel
generators (EDGs) and decrease the 24-month test load for the Unit 2
EDGs in TS Surveillance Requirement 3.8.1.9.
Date of issuance: June 21, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 196, 185.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 2009 (74 FR
4774).
The supplemental letters contained clarifying information, did not
change the initial no significant hazards consideration determination,
and did not expand the scope of the original Federal Register notice.
[[Page 39984]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 21, 2010.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: July 30, 2009.
Brief description of amendment: The amendment relocates the
Technical Specification (TS) surveillance requirement for the reactor
recirculation system motor-generator set scoop tube stop settings to
the Technical Requirements Manual.
Date of issuance: June 28, 2010.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 181.
Facility Operating License No. NPF-57: The amendment revised the
TSs and the License.
Date of initial notice in Federal Register: October 6, 2009 (74 FR
51333).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 2010.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: August 10, 2009.
Brief description of amendments: The amendment modified Technical
Specification 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to allow a
7-day Completion Time for the turbine-driven AFW pump if the
inoperability of the pump occurs in MODE 3 following a refueling
outage, and if MODE 2 has not been entered. This change is consistent
with the U.S. Nuclear Regulatory Commission-approved Technical
Specification Task Force (TSTF) traveler, TSTF-340,
Revision 3.
Date of issuance: June 30, 2010.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 2-223; Unit 3-216.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: November 17, 2009 (74
FR 59263).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 30, 2010.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: May 4, 2009.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.7.3, ``Main Feedwater Isolation Valves (MFIVs) and
Main Feedwater Regulating Valves (MFRVs) and Main Feedwater Regulating
Valve Bypass Valves (MFRVBVs),'' so that the Limiting Condition for
Operation (LCO) and Applicability more accurately reflect the
conditions for when the LCO should be applicable and more effectively
provide appropriate exceptions to the Applicability for certain valve
configurations. The amendment also changed the title of TS 3.7.3 to
``Main Feedwater Isolation Valves (MFIVs), Main Feedwater Regulating
Valves (MFRVs), and Main Feedwater Regulating Valve Bypass Valves
(MFRVBVs),'' and the associated page header to ``MFIVs, MFRVs, and
MFRVBVs.'' In addition, the amendment revised footnotes to TS 3.3.2,
``Engineered Safety Feature Actuation System (ESFAS) Instrumentation,''
Table 3.3.2-1, in order to improve application of existing notes and/or
incorporate more appropriate notes.
Date of issuance: June 29, 2010.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 198.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 25, 2009 (74 FR
42932).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 2010.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: July 10, 2009.
Brief description of amendment: The amendment deletes the Technical
Specification (TS) requirements for the containment hydrogen
recombiners and relaxes the requirements for hydrogen and oxygen
monitors. The TS changes support implementation of the revisions to
Title 10 of the Code of Federal Regulations (10 CFR) Sec. 50.44,
``Combustible gas control for nuclear power reactors,'' that became
effective on October 16, 2003. The changes are consistent with Revision
1 of the NRC-approved Industry/Technical Specification Task Force
(TSTF) Standard Technical Specification Change Traveler, TSTF-447,
``Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen
Monitors.'' This operating license improvement was made available by
the NRC on September 25, 2003 (68 FR 55416), as part of the
consolidated line item improvement process. In addition, the amendment
corrected four typographical errors in the TSs.
Date of issuance: June 29, 2010.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 199.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 25, 2009 (74 FR
42934).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 2010.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: June 1, 2009, as supplemented by
letters dated August 27, 2009, and March 4, 2010.
Brief description of amendment: The amendment revised the Limiting
Condition for Operation (LCO) Applicability Note for Technical
Specification (TS) 3.3.9, ``Boron Dilution Mitigation System (BDMS).''
The LCO Applicability Note was revised to clarify the situations during
which the BDMS signal may be blocked in MODES 2 and 3.
Date of issuance: June 29, 2010.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 200.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 25, 2009 (74 FR
[[Page 39985]]
42933). The supplemental letters dated August 27, 2009, and March 4,
2010, 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 June 29, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 1st day of July 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-16879 Filed 7-12-10; 8:45 am]
BILLING CODE 7590-01-P