[Federal Register Volume 78, Number 14 (Tuesday, January 22, 2013)]
[Notices]
[Pages 4469-4476]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-01010]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0012]
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 December 27, 2012 to January 9, 2013. The
last biweekly notice was published on January 8, 2013 (78 FR 1267).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and are publically available,
by searching on http://www.regulations.gov under Docket ID NRC-2013-
0012. 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-2013-0012. 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-2013-0012 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document by any of the following
methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0012.
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 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-2013-0012 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 that that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely 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 that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely 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 section 50.92 of Title 10 of the Code
of Federal Regulations (10 CFR), 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
[[Page 4470]]
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 held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 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,
[[Page 4471]]
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 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. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the 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].
Detroit Edision, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: November 13, 2012.
Description of amendment request: The proposed amendment would
modify Technical Specification requirements to operate ventilation
systems with charcoal filters for 10 hours each in accordance with
Technical Specifications Task Force (TSTF)-522, Revision 0, ``Revise
Ventilation System Surveillance Requirements to Operate for 10 hours
per Month.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the SGT System and CREF System equipped with
electric heaters for a continuous 10 hour period every 31 days with
a requirement to operate the systems for 15 continuous minutes with
heaters operating.
These systems are not accident initiators and therefore, these
changes do not involve a significant increase in the probability of
an accident. The proposed system and filter testing changes are
consistent with current regulatory guidance for these systems and
will continue to assure that these systems perform their design
function which may include mitigating accidents. Thus the change
does not involve a significant increase in the consequences of an
accident.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the SGT System and CREF System equipped with
electric heaters for a continuous 10 hour period every 31 days with
a requirement to operate the systems for 15 continuous minutes with
heaters operating.
[[Page 4472]]
The change proposed for these ventilation systems does not
change any system operations or maintenance activities. Testing
requirements will be revised and will continue to demonstrate that
the Limiting Conditions for Operation are met and the system
components are capable of performing their intended safety
functions. The change does not create new failure modes or
mechanisms and no new accident precursors are generated.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the SGT System and CREF System equipped with
electric heaters for a continuous 10 hour period every 31 days with
a requirement to operate the systems for 15 continuous minutes with
heaters operating.
The design basis for the ventilation systems' heaters is to heat
the incoming air which reduces the relative humidity. The heater
testing change proposed will continue to demonstrate that the
heaters are capable of heating the air and will perform their design
function. The proposed change is consistent with regulatory
guidance.
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: Bruce R. Masters, DTE Energy, General
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
NRC Branch Chief: Robert D. Carlson.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone
Power Station, Unit 2, New London County, Connecticut
Date of amendment request: December 17, 2012.
Description of amendment request: The proposed amendment would
revise the Millstone Power Station, Unit 2 (MPS2) Technical
Specification (TS) Surveillance Requirement 4.4.3.2 to remove the
requirement to perform the quarterly surveillance for a pressurizer
power-operated relief valve (PORV) block valve that is being maintained
closed in accordance with TS 3.4.3 Action a. The proposed change is
consistent with the requirements of the standard Technical
Specification for Combustion Engineering plants (NUREG-1432).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), of Title 10 of the Code
of Federal Regulations (10 CFR), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
Criterion 1
Will operation of the facility in accordance with the proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The block valve for the pressurizer PORV is not a potential
accident initiator. Therefore, not requiring a surveillance of the
block valve while it is being used to isolate its associated PORV
will not increase the probability of an accident previously
evaluated. Not requiring the surveillance of the block valve may
slightly reduce the probability of a loss of coolant accident from a
stuck open PORV since it will eliminate the challenge to the PORV
from the pressure transient that results from cycling the block
valve.
The PORVs or the PORV block valves are not credited in the MPS2
Final Safety Analysis Report (FSAR), Chapter 14, ``Safety
Analysis,'' for event mitigation. If pressurizer spray is not
available or is not effective, either one or the two pressurizer
PORVs may be manually actuated to depressurize the RCS in response
to certain transients. Not performing the surveillance on the block
valve is not relevant to the primary system for depressurizing the
RCS (pressurizer spray). The block valves have been demonstrated by
operating experience to be reliable and are also subject to the
motor-operated valve testing program. Consequently, the proposed
change does not significantly reduce the confidence that the block
valve can be opened to permit manual actuation of the PORV to
depressurize the RCS.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2
Will operation of the facility in accordance with the proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated?
Response: No.
The proposed change only affects the performance of the
surveillance test for the block valve and does not involve any
physical alteration of plant equipment or introduce any operating
configurations not previously evaluated. The pressurizer PORV block
valves provide isolation for a postulated stuck-open or leaking
PORV. Isolation is satisfied with the block valve closed in
accordance with SR 4.4.3.2. PORV block valve closure is not credited
in FSAR Chapter 14 for inadvertent opening of the PORV event
mitigation.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
Criterion 3
Will operation of the facility in accordance with the proposed
change involve a significant reduction in the margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
These barriers are not significantly affected by the changes
proposed herein. The margin of safety is established through the
design of the plant structures, systems, and components, the
parameters within which the plant is operated, and the establishment
of setpoints for the actuation of equipment relied upon to respond
to an event, and thereby protect the fission product barriers. The
proposed change to the surveillance requirement for the presurrizer
PORV block valve does not affect the assumptions in any accident
analysis.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: George A. Wilson.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County,
Texas
Date of amendment request: October 2, 2012.
Brief description of amendments: The amendments would revise
Technical Specification (TS) 3.3.1, ``Reactor Trip System (RTS)
Instrumentation,'' and TS 3.3.2, ``Engineered Safety Feature Actuation
System (ESFAS) Instrumentation,'' to relocate the TS requirements for
the following instruments to the Technical Requirements Manual (TRM), a
licensee-controlled document, under 10 CFR 50.59:
Pressurizer Water level--High (RTS Function No. 9)
Trip of all Main Feedwater Pumps (ESFAS Function No. 6.g)
ESFAS Interlock, Reactor Trip, P-4 (ESFAS Function No.
8.a)
The proposed changes would relocate the TS requirements in their
entirety and not result in deletion or alteration of any RTS or ESFAS
requirements. The proposed relocation of the TS requirements for these
RTS and ESFAS instrument Functions is based on the
[[Page 4473]]
application of the TS criteria of 10 CFR 50.36(c)(2)(ii).
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.
The proposed change to the TS does not affect the initiators of
any analyzed accident. In addition, operation in accordance with the
proposed TS change will continue to ensure that the previously
evaluated accidents will be mitigated as analyzed. Thus, the
proposed change does not adversely affect the design function or
operation of any structures, systems, and components important to
safety.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
The proposed change does not create any new failure modes for
existing equipment or any new limiting single failures. Additionally
the proposed change does not involve a change in the methods
governing normal plant operation and all safety functions will
continue to perform as previously assumed in accident analyses.
Thus, the proposed change does not adversely affect the design
function or operation of any structures, systems, and components
important to safety.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not adversely affect the operation of
plant equipment or the function of equipment assumed in the accident
analyses. The proposed changes to the RTS and ESFAS TS requirements
do not change the RTS or ESFAS design and capability to perform the
required safety functions consistent with the assumptions of the
applicable safety analyses. In addition, operation in accordance
with the proposed TS change will continue to ensure that the
previously evaluated accidents will be mitigated as analyzed.
Therefore, the proposed change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
NextEra Energy Seabrook, LLC Docket No. 50-443, Seabrook Station,
Unit 1, Rockingham County, New Hampshire
Date of amendment request: December 20, 2012.
Description of amendment request: The proposed amendment will
revise the Seabrook Technical Specifications (TS) TS 6.7.6.m, ``Reactor
Coolant Pump Flywheel Inspection Program.'' The proposed amendment will
extend the reactor coolant pump (RCP) motor flywheel examination
frequency from the currently approved 10-year inspection interval, to
an interval not to exceed 20 years. The changes are consistent with
Industry/Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-421, ``Revision to RCP Flywheel
Inspection Program (WCAP-15666).'' The availability of this TS
improvement was announced in the Federal Register on October 22, 2003,
as part of the consolidated line item improvement process (CLIIP).
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 (NSHC) through incorporation by reference of the NSHC
published in the Federal Register Notice dated June 24, 2003 (68 FR
37590), which is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines contained in RG 1.174 (<1.0E-6 per year). Moreover,
considering the uncertainties involved in this evaluation, the risk
associated with the postulated failure of an RCP motor flywheel is
significantly low. Even if all four RCP motor flywheels are
considered in the bounding plant configuration case, the risk is
still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
The proposed change in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the analysis and, based on this review,
it appears that the three standards of
[[Page 4474]]
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves NSHC.
Attorney for licensee: James Petro, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Meena Khanna.
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating
Station, Unit 1, Salem County, New Jersey
Date of amendment request: May 8, 2012.
Description of amendment request: The proposed amendment would
revise Salem Unit 1 Technical Specification (TS) 6.8.4.i, ``Steam
Generator (SG) Program,'' to permanently exclude portions of the tube
below the top of the steam generator tubesheet from periodic steam
generator tube inspections. In addition, this amendment proposes to
revise TS 6.9.1.10, ``Steam Generator Tube Inspection Report,'' to
provide permanent reporting requirements that have been previously
established on an interim basis.
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, with the NRC staff edits in
square brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
that alters the steam generator inspection criteria does not have a
detrimental impact on the integrity of any plant structure, system,
or component that initiates an analyzed event. The proposed change
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident.
Of the applicable accidents previously evaluated, the limiting
transients with consideration to the proposed change to the steam
generator tube inspection and repair criteria are the steam
generator tube rupture (SGTR) event, the steam line break (SLB) and
the feedline break (FLB) postulated accidents.
Addressing the SGTR event, the required structural integrity
margins of the steam generator tubes and the tube-to-tubesheet joint
over the H* distance will be maintained. Tube rupture in tubes with
cracks within the tubesheet is precluded by the presence of the
tubesheet and constraint provided by the tube-to-tubesheet joint.
Tube burst cannot occur within the thickness of the tubesheet. The
tube-to-tubesheet joint constraint results from the hydraulic
expansion process, thermal expansion mismatch between the tube and
tubesheet, from the differential pressure between the primary and
secondary side, and tubesheet deflection. The structural margins
against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases
for Plugging Degraded PWR [pressurized-water reactor] Steam
Generator Tubes,'' and TS 6.8.4.i are maintained for both normal and
postulated accident conditions.
The proposed change has no impact on the structural or leakage
integrity of the portion of the tube outside of the tubesheet. The
proposed change maintains structural and leakage integrity of the
steam generator tubes consistent with the performance criteria in TS
6.8.4.i. Therefore, the proposed change results in no significant
increase in the probability of the occurrence of a SGTR accident.
At normal operating pressures, leakage from tube degradation
below the proposed limited inspection depth is limited by the tube-
to-tubesheet joint. Consequently, negligible normal operating
leakage is expected from degradation below the inspected depth
within the tubesheet region. The consequences of an SGTR event are
not affected by the primary to secondary leakage flow during the
event as primary to secondary leakage flow through a postulated tube
that has been pulled out of the tubesheet is essentially equivalent
to a severed tube. Therefore, the proposed changes do not result in
a significant increase in the consequences of a SGTR.
The consequences of a SLB or FLB are also not significantly
affected by the proposed changes. The leakage analysis shows that
the primary-to-secondary leakage during a SLB/FLB event would be
less than or equal to that assumed in the Updated Safety Analysis
Report.
Primary-to-secondary leakage from tube degradation in the
tubesheet area during the limiting accidents (i.e., SLB/FLB) is
limited by flow restrictions. These restrictions result from the
crack and tube-to-tubesheet contact pressures that provide a
restricted leakage path above the indications and also limit the
degree of potential crack face opening as compared to free span
indications.
The leakage factor for Salem Unit 1, for a postulated SLB/FLB,
has been calculated as 2.16. Specifically, for the condition
monitoring (CM) assessment, the component of leakage from the prior
cycle from below the H* distance will be multiplied by a factor of
2.16 and added to the total leakage from any other source and
compared to the allowable accident induced leakage limit. For the
operational assessment (OA), the difference in the leakage between
the allowable leakage and the accident induced leakage from sources
other than the tubesheet expansion region will be divided by 2.16
and compared to the observed operational leakage.
The probability of an SLB/FLB is unaffected by the potential
failure of a steam generator tube as the failure of the tube is not
an initiator for an SLB/FLB event. SLB/FLB leakage is limited by
leakage flow restrictions resulting from the leakage path above
potential cracks through the tube-to-tubesheet crevice. The leak
rate during all postulated accident conditions that model primary-
to-secondary leakage (including locked rotor and control rod
ejection) has been shown to remain within the accident analysis
assumptions for all axial and or circumferentially orientated cracks
occurring 15.21 inches below the top of the tubesheet. The accident
analysis calculations have an assumption of 0.6 gpm [gallons per
minute] at room temperature (gpmRT) primary-to-secondary leakage in
a single SG and 1 gpm at room temperature (gpmRT) total primary-to-
secondary leakage for all SGs. This apportioned primary-to-secondary
leakage is used in the Main Steam Line Break and Locked Rotor
accidents. Primary-to-secondary leakage of 1 gpm at room temperature
(gpmRT) from all SGs, conservatively modeled to be released from a
single location to maximize control room dose consequences, is used
in the Control Rod Ejection (CRE) accident. The TS operational leak
rate limit is 150 gallons per day (gpd) (0.104 gpmRT). The maximum
accident leak rate ratio for Salem Unit 1 is 2.16 (Revised Table 9-
7, Reference 15, [of the licensee's amendment request dated May 8,
2012]). Consequently, this results in significant margin between the
conservatively estimated accident leakage and the allowable accident
leakage.
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 proposed change alters the steam generator inspection and
reporting criteria. It does not introduce any new equipment, create
new failure modes for existing equipment, or create any new limiting
single failures. Plant operation will not be altered, and safety
functions will continue to perform as previously assumed in accident
analyses.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed change alters the steam generator inspection and
reporting criteria. It maintains the required structural margins of
the steam generator tubes for both normal and accident conditions.
NEI 97-06 and RG 1.121, are used as the bases in the development of
the limited tubesheet inspection depth methodology for determining
that steam generator tube integrity considerations are maintained
within acceptable limits. RG 1.121 describes a method acceptable to
the NRC for meeting GDC [General Design Criteria] 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
[[Page 4475]]
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, the H* Analysis documented in
Section 3, [of the licensee's amendment request dated May 8, 2012,]
defines a length of degradation-free expanded tubing that provides
the necessary resistance to tube pullout due to the pressure induced
forces, with applicable safety factors applied. Application of the
limited hot and cold leg tubesheet inspection criteria will preclude
unacceptable primary to secondary leakage during all plant
conditions. The methodology for determining leakage provides for
large margins between calculated and actual leakage values in the
proposed limited tubesheet inspection depth criteria.
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, and with the changes noted above in square brackets, 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Meena K. Khanna.
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 NRC'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].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone
Power Station, Unit 2, New London County, Connecticut
Date of amendment request: July 31, 2012.
Description of amendment request: The proposed amendment would
revise the Millstone Power Station, Unit 2 Technical Specification
requirements regarding steam generator tube inspections and reporting
as described in TSTF-510, Revision 2, ``Revision to Steam Generator
Program Inspection Frequencies and Tube Sample Selection;'' however,
Dominion Nuclear Connecticut, Inc. is proposing minor variations and
deviations from TSTF-510.
Date of issuance: January 4, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 312.
Renewed Facility Operating License No. DPR-65: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: September 4, 2012 (77
FR 53926).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 4, 2013.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana
Parish, Louisiana
Entergy Nuclear Operations, Inc., Docket Nos. 50-155 and 72-043
(ISFSI), Big Rock Point Plant (Big Rock), Charlevoix County, Michigan
Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247 and
50-286, Indian Point Nuclear Generating Units 1, 2 and 3 (IP1, IP2, and
IP3), Westchester County, New York
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego County, New York
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades
Nuclear Plant (Palisades), Van Buren County, Michigan
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim
Nuclear Power Station (Pilgrim), Plymouth County, Massachusetts
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station (VY),
Vernon, Vermont
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2 (ANO1 and ANO2), Pope County, Arkansas
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 (GGNS), Claiborne
County, Mississippi
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3 (Waterford), St. Charles Parish, Louisiana
Date of application for amendment: December 13, 2011, as
supplemented by letters dated May 21, and November 20, 2012.
Brief description of amendment: The amendments approved changes to
the Quality Assurance Program Manual (QAPM) and Technical
Specifications (TSs) for the above specified plants. The proposed
changes standardize unit staff qualification requirements for the
Entergy fleet. Certain changes to the QAPM are a reduction in
commitment and, in accordance with 10 CFR 50.54(a)(4), NRC approval is
required prior to implementation.
Date of issuance: December 28, 2012.
Effective date: As of the date of issuance and shall be implemented
120 days from the date of issuance.
Amendment Nos.: ANO1--248; ANO2--296; FitzPatrick--304; GGNS--193;
IP2--271; IP3--248; Palisades--
[[Page 4476]]
249; Pilgrim--239; RBS--178; VY--253; and Waterford--240.
Facility Operating License Nos. DPR-51, NPF-6, NPF-29, NPF-47, NPF-
38, DPR-59, DPR-35, DPR-26, DPR-64, DPR-20, and DPR-28: The amendments
revise the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 20, 2012 (77 FR
16274). The supplemental letters dated May 21 and November 20, 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 amendments is contained
in a Safety Evaluation dated December 28, 2012.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit 1, Washington County, Nebraska
Date of amendment request: December 23, 2011, as supplement by
letter dated June 18, 2012.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to incorporate a new Radial Peaking Factor
definition and to clarify Limiting Condition for Operation 2.10.2(6),
``Shutdown CEA [Control Element Assembly] Insertion Limit During Power
Operation.'' Specifically, the amendment removed requirements for, and
references to, the ``Unrodded Integrated Radial Peaking Factor.'' The
amendment also added a definition of, and references to, the ``Maximum
Radial Peaking Factor (FR\T\).'' Additional clarifications
and editorial changes were made to TS 2.10, ``Reactor Core.''
Date of issuance: December 31, 2012.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 269.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: April 17, 2012 (74 FR
22816). The supplemental letter dated June 18, 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 December 31, 2012.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo
County, California
Date of application for amendment: January 5, 2012.
Brief description of amendment: The amendments revised the Diablo
Canyon Power Plant, Units 1 and 2, Final Safety Analysis Report Update
Section 4.3.2.2, ``Power Distribution,'' to allow the use of the
Westinghouse Electric Company LLC's Best Estimate Analyzer for the Core
Operations-Nuclear (BEACON) Power Distribution Monitoring System
methodology as described in WCAP-12472-P-A, Addendum 1-A, ``BEACON Core
Monitoring and Operation Support System,'' January 2000.
Date of issuance: January 9, 2013.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance. Implementation of the
amendments shall also include revision of the Final Safety Analysis
Report Update as described in the licensee's letter dated January 5,
2012.
Amendment Nos.: Unit 1--214; Unit 2--216.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: May 15, 2012 (77 FR
28633).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 9, 2013.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: January 18, 2012.
Brief description of amendment request: The amendment revises
Technical Specification (TS) Surveillance Requirements 3.4.11.1 and
3.4.11.4 by removing requirements no longer applicable to Joseph M.
Farley Nuclear Plant, Unit 2.
Date of issuance: December 27, 2012.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 186.
Facility Operating License No. NPF-8:
Date of initial notice in Federal Register: October 2, 2012 (77 FR
60152).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 27, 2012.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 11th day of January 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-01010 Filed 1-18-13; 8:45 am]
BILLING CODE 7590-01-P