[Federal Register Volume 78, Number 228 (Tuesday, November 26, 2013)]
[Notices]
[Pages 70589-70596]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-28225]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0257]
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 (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to 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 October 31, 2013 to November 13, 2013. The
last biweekly notice was published on November 12, 2013 (78 FR 67402).
ADDRESSES: You may submit comment by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0257. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact
the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: 3WFN, 06-44M, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
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-2013-0257 when contacting the NRC
about the availability of information regarding this document. You may
access publicly-available information related to this action by the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0257.
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]. The ADAMS accession number
for each document referenced in this notice (if that document is
available in ADAMS) is provided the first time that a document is
referenced.
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-2013-0257 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 you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering 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 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
[[Page 70590]]
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 ``Agency
Rules of Practice and Procedure'' 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's E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then
[[Page 70591]]
submit a request for hearing or petition for leave to intervene.
Submissions should be in Portable Document Format (PDF) in accordance
with the NRC guidance available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered
complete at the time the documents are submitted through the NRC's E-
Filing system. To be timely, an electronic filing must be submitted to
the E-Filing system no later than 11:59 p.m. Eastern Time on the due
date. Upon receipt of a transmission, the E-Filing system time-stamps
the document and sends the submitter an email notice confirming receipt
of the document. The E-Filing system also distributes an email notice
that provides access to the document to the NRC's Office of the General
Counsel and any others who have advised the Office of the Secretary
that they wish to participate in the proceeding, so that the filer need
not serve the documents on those participants separately. Therefore,
applicants and other participants (or their counsel or representative)
must apply for and receive a digital ID certificate before a hearing
request/petition to intervene is filed so that they can obtain access
to the document via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email to
[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.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. 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 three factors in 10
CFR 2.309(c)(1)(i)(iii).
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].
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: October 30, 2012, as supplemented on
January 21, June 11, September 3, and October 21, 2013.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) to allow operation of a
reverse osmosis system during normal plant operation to purify the
water in the borated water storage tanks and the spent fuel pools.
Automatic isolation valves would be installed in the Spent Fuel Pool
Cooling (SFPC) system upstream of the Reverse Osmosis (RO) system
borated water storage tank suction connections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), in its supplemental
letter dated October 21, 2013, the licensee provided a revised 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 requests Nuclear Regulatory Commission (NRC)
approval of design features and controls that will be used to ensure
that unisolating the SFPC Purification System and the Reverse
Osmosis (RO) System during Unit operations does not significantly
impact the Borated Water Storage Tank (BWST) or other plant
equipment and that periodic limited operation of the RO System when
aligned to a SFP during Unit operation does not significantly impact
the Spent Fuel Pool (SFP) function or other plant equipment. The
proposed change also requests NRC to approve proposed Technical
Specification (TS) requirements that will impose operating
restrictions and isolation requirements for the SFPC Purification
System and the RO System.
The new high energy piping and non-seismic piping being
installed for the RO System is non-QA1 and is postulated to fail.
Adequate measures have been provided to isolate the flood source
(BWST or SFP) prior to affecting SSCs important to safety.
The BWST will be automatically isolated prior to going below the
TS water volume requirement. For the SFP, the suction to the RO
system is above the required TS water level, therefore, the design
ensures the required TS water level is maintained.
Procedural controls will ensure that the boron concentration
does not go below the TS limit as a result of water returned from
the RO System with lower boron concentration. Thus, no adverse
effects from decreased boron concentration will occur.
The RO System takes suction from the top of the SFP to protect
SFP inventory. Plant
[[Page 70592]]
procedures will prohibit the use of the RO System for the Units 1
and 2 SFP during the time period directly after an outage that
requires the Units 1 and 2 SFP level to be maintained higher than
the TS Limiting Condition for Operation (LCO) 3.7.11 level
requirement. The higher level is required to support TS LCO 3.10.1
requirements for Standby Shutdown Facility (SSF) Reactor Coolant
(RC) Makeup System operability (due to the additional decay heat
from the recently offloaded spent fuel). Plant procedures will also
specify the siphon be broken during this time period so the SFP
water above the RO suction point cannot be siphoned off if the RO
piping breaks. The proposed change does not impact the fuel
assemblies, the movement of fuel, or the movement of fuel shipping
casks. The SFP boron concentration, level, and temperature limits
will not be outside of required parameters due to restrictions/
requirements on the system's operation. In addition, the proposed
new Technical Specification will require the siphon be broken during
movement of irradiated fuel assemblies in the SFP or movement of a
cask over the SFP. Therefore, RO System operation cannot occur
during these activities, effectively eliminating a Fuel Handling
Accident (FHA) from occurring while the RO System is in operation.
The BWST is used for mitigation of Steam Generator Tube Rupture
(SGTR), Main Steam Line Break (MSLB), and Loss of Coolant Accidents
(LOCAs). The SGTR and MSLB are bounded by the small break (SBLOCA)
analyses with respect to the performance requirements for the High
Pressure Injection (HPI) System. In the normal mode of Unit
operation, the BWST is not an accident initiator. The SFP is
evaluated to maintain acceptable criticality margin for all abnormal
and accident conditions including FHAs and cask drop accidents. Both
the BWST and SFP are specified by TS requirements to have minimum
levels/volumes and boron concentrations. The BWST also has TS
requirements for temperature. Prior to RO System operation,
procedures will require the minimum required initial boron
concentration and initial level/volume to be adjusted. Additionally,
they will require the RO System operation to be restricted to a
specified maximum time period before readjusting volume and boron
concentration prior to another RO session. This ensures that the TS
specified boron concentration and level/volume limits for both the
SFP and the BWST are not exceeded during RO System operation. Thus,
the design functions of the BWST and the SFP will continue to be met
during RO System operation.
The proposed TS will require the RO system to be isolated (by
breaking the siphon) from the SFPs during fuel handling activities
and will require the automatic isolation valves between the BWST and
the SFPC Purification System, upstream of the branch line to the RO
System branch line, be OPERABLE in MODES 1, 2, 3, and 4. The TS will
also require manual valves in branch lines upstream of the SFPC
Purification System automatic isolation valves to be closed and meet
Inservice Testing (IST) Program leakage requirements.
The additional controls imposed by the proposed Technical
Specifications will provide additional assurance that isolation
valves and operating restrictions credited to eliminate the need to
analyze new release pathways will be in place.
Therefore, allowing the SFPC Purification System and the RO
System to be unisolated during Unit operation do not significantly
increase the probability or consequences of any accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The RO System adds non-seismic piping in the Auxiliary Building.
However, the break of a single non-seismic pipe in the Auxiliary
Building has already been postulated as an event in the licensing
basis. The RO System also does not create the possibility of a
seismic event concurrent with a LOCA since a seismic event is a
natural phenomenon event. The RO System does not adversely affect
the Reactor Coolant System pressure boundary.
Duke Energy also evaluated potential releases of radioactive
liquid to the environment. Design features, controls imposed by the
proposed Technical Specification, and procedural controls will
preclude release of radioactive materials outside the Auxiliary
Building by ensuring the SFPC Purification System and the RO System
will be isolated when required.
The SFP suction line is designed such that the SFP water level
will not go below TS required levels, thus the fuel assemblies will
have the TS required water level over them. Procedural controls will
restrict the use of the RO System and require breaking vacuum on the
Units 1 and 2 SFP suction line when the SSF conditions require the
SFP level be raised to support SSF RC Makeup System operability.
Thus, the SFP water level will not be reduced below required water
levels for these conditions. RO System operating restrictions will
prevent reducing the SFP boron concentration below TS limits.
Since the BWST and SFP already have TS boron concentration and
level/volume requirements and the RO System will be automatically
isolated, the mitigation of a LOCA or FHA does not result in an
increase in dose consequence. The design basis LOCA analysis for
Oconee assumes 5 gpm back-leakage from the Reactor Building sump to
the BWST. The automatic isolation valves will isolate on a BWST
level prior to swapover to the recirculation phase and prior to
going below the actual TS water level. The proposed TS will require
the RO system to be isolated (by breaking the siphon) from the SFPs
prior to movement of irradiated fuel assemblies in the SFP or
movement of cask over the SFP and will require the automatic
isolation valves between the BWST and RO System to be OPERABLE in
MODES 1, 2, 3, and 4.
The additional controls imposed by the proposed Technical
Specifications will provide additional assurance that isolation
valves and operating restrictions credited to eliminate the need to
analyze new release pathways (introduced by allowing the SFPC
Purification System and the RO system to be unisolated during Unit
operation) will be in place.
Therefore, operation of these systems unisolated will not create
the possibility of a new or different kind of accident from any kind
of accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Duke Energy evaluated the impact of allowing the SFPC
Purification System and the RO System to be unisolated during Unit
operation on SSCs important to safety and determined that the
proposed TS controls and procedural controls will ensure that TS
limits for SFP and BWST volume, temperature, and boron concentration
will continue to be met. For the BWST, these controls will ensure
the TS minimum BWST boron concentration and level are available to
mitigate the consequences of a small break LOCA or a large break
LOCA. For the SFP, these controls ensure the assumptions of the fuel
handling and cask drop accident analyses are preserved. The proposed
change does not significantly impact the condition or performance of
SSCs relied upon for accident mitigation. This change does not alter
the existing TS allowable values or analytical limits. The existing
operating margin between Unit conditions and actual Unit setpoints
is not significantly reduced due to these changes. The assumptions
and results in any safety analyses are not impacted. Therefore,
operation of the RO System during Unit operation 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: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: June 7, 2013.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) 1.1 ``Definitions,'' for
Shutdown Margin (SDM), to require calculation of the SDM at a reactor
moderator temperature of 68 [deg]F or a higher temperature that is
determined to represent the most reactive state throughout the
operating cycle of the reactor. This change is needed to address new
Boiling Water Reactor (BWR) fuel designs which may be more reactive at
shutdown temperatures above 68 [deg]F.
[[Page 70593]]
The NRC staff announced the availability of Technical
Specifications (TSs) Task Force (TSTF) Traveler TSTF-535, Revision 0,
``Revise Shutdown Margin Definition to Address Advanced Fuel Designs.''
The TSTF-535, Revision 0 provides guidance for plant-specific adoption
of changes needed to address BWR fuel designs which may be more
reactive at shutdown temperatures greater than 68 [deg]F, using the
agency's Consolidated Line Item Improvement Process'' (CLIIP). The
availability and the model safety evaluation of TSTF-535, Revision 0,
was provided under ADAMS Accession No. ML12355A772, and published in
the Federal Register dated November 19, 2012 (77 FR 69507).
The licensee has reviewed the information provided by the NRC staff
in TSTF-535, and the model safety evaluation, as announced in the
Federal Register (FR) Notice of availability. The licensee concluded
that the justification presented in the FR Notice of availability of
TSTF-535, Revision 0 and the model safety evaluation, prepared by the
NRC staff, is applicable to the James A. FitzPatrick Nuclear Power
Plant and justifies the current request for amendment to TS 1.1,
``Definitions'' for SDM.
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 revises the definition of SDM. SDM is not an
initiator to any accident previously evaluated. Accordingly, the
proposed change to the definition of SDM has no effect on the
probability of any accident previously evaluated. SDM is an
assumption in the analysis of some previously evaluated accidents
and inadequate SDM could lead to an increase in consequences for
those accidents. However, the proposed change revises the SDM
definition to ensure that the correct SDM is determined for all fuel
types at all times during the fuel cycle. As a result, the proposed
change does not adversely affect the consequences of any 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 accident from any accident previously
evaluated?
Response: No.
The proposed change revises the definition of SDM. The change
does not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. The change does not alter
assumptions made in the safety analysis regarding SDM.
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 definition of SDM. The proposed
change does not alter the manner in which safety limits, limiting
safety system settings or limiting conditions for operation are
determined. The proposed change ensures that the SDM assumed in
determining safety limits, limiting safety system settings or
limiting conditions for operation is correct for all BWR fuel types
at all times during the fuel cycle.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: R. Beall.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: April 24, 2012, as supplemented by
letters dated July 12 and August 23, 2012, and January 14, February 12,
March 13, and June 13, 2013.
Description of amendment request: The proposed amendment would
adopt National Fire Protection Association (NFPA) 805, ``Performance-
Based Standard for Fire Protection for Light Water Reactor Generating
Plants'' (2001 Edition). Implementation of the regulatory actions
presented in the attachments to the license amendment request will
enable Cooper Nuclear Station to adopt a new fire protection licensing
basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR
50.48(c), and the guidance in Regulatory Guide (RG) 1.205, Revision 1.
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.
Operation of the Cooper Nuclear Station (CNS) in accordance with
the proposed amendment does not result in a significant increase in
the probability or consequences of accidents previously evaluated.
The proposed amendment does not affect accident initiators or
precursors as described in the CNS Updated Safety Analysis Report
(USAR), nor does it adversely alter design assumptions, conditions,
or configurations of the facility, and it does not adversely impact
the ability of structures, systems, or components (SSCs) to perform
their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits. The proposed
changes do not affect the way in which safety-related systems
perform their functions as required by the accident analysis. The
SSCs required to safely shut down the reactor and to maintain it in
a safe shutdown condition will remain capable of performing their
design functions.
The purpose of this amendment is to permit CNS to adopt a new
risk-informed, performance-based fire protection licensing basis
that complies with the requirements in 10 CFR 50.48(a) and 10 CFR
50.48(c), as well as the guidance contained in Regulatory Guide (RG)
1.205. The NRC considers that NFPA 805 provides an acceptable
methodology and performance criteria for licensees to identify fire
protection requirements that are an acceptable alternative to the 10
CFR part 50, appendix R, fire protection features (69 FR 33536; June
16, 2004). Engineering analyses, which may include engineering
evaluations, probabilistic risk assessments, and fire modeling
calculations, have been performed to demonstrate that the
performance-based requirements of NFPA 805 have been met.
NFPA 805, taken as a whole, provides an acceptable alternative
for satisfying General Design Criterion 3 (GDC 3) of appendix A to
10 CFR part 50. It meets the underlying intent of the NRC's existing
fire protection regulations and guidance, and achieves defense-in-
depth along with the goals, performance objectives, and performance
criteria specified in NFPA 805, Chapter 1. In addition, if there are
any increases in core damage frequency (CDF) or risk as a result of
the transition to NFPA 805, the increase will be small, governed by
the delta risk requirements of NFPA 805, and consistent with the
intent of the Commission's Safety Goal Policy.
Based on the above, the implementation of this amendment to
transition the Fire Protection Plan (FPP) at CNS to one based on
NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a
significant increase in the probability of any accident previously
evaluated. In addition, all equipment required to mitigate an
accident remains capable of performing the assumed function.
[[Page 70594]]
Therefore, the consequences of any accident previously evaluated
are not significantly increased with the implementation of this
License Amendment Request.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Operation of CNS in accordance with the proposed license
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated. Any scenario or
previously analyzed accident with offsite dose consequences was
included in the evaluation of design basis accidents (DBA)
documented in the USAR as a part of the transition to NFPA 805. The
proposed amendment does not impact these accident analyses. The
proposed change does not alter the requirements or functions for
systems required during accident conditions, nor does it alter the
required mitigation capability of the fire protection program, or
its functioning during accident conditions as assumed in the
licensing basis analyses and/or DBA radiological consequences
evaluations.
The proposed amendment does not adversely affect accident
initiators nor alter design assumptions, or conditions of the
facility. The proposed amendment does not adversely affect the
ability of SSCs to perform their design function. SSCs required to
maintain the unit in a safe and stable condition remain capable of
performing their design functions.
The purpose of the proposed amendment is to permit CNS to adopt
a new fire protection licensing basis which complies with the
requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance
in Revision 1 of RG 1.205. As indicated in the Statements of
Consideration, the NRC considers that NFPA 805 provides an
acceptable methodology and performance criteria for licensees to
identify fire protection systems and features that are an acceptable
alternative to the 10 CFR part 50, appendix R fire protection
features.
The requirements in NFPA 805 address only fire protection and
the impacts of fire effects on the plant have been evaluated. The
proposed fire protection program changes do not involve new failure
mechanisms or malfunctions that could initiate a new or different
kind of accident beyond those already analyzed in the USAR. Based on
this, as well as the discussion above, the implementation of this
amendment to transition the FPP at CNS to one based on NFPA 805, in
accordance with 10 CFR 50.48(c), 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.
Operation of CNS in accordance with the proposed license
amendment does not involve a significant reduction in a margin of
safety. The transition to a new risk-informed, performance-based
fire protection licensing basis that complies with the requirements
in 10 CFR 50.48(a) and 10 CFR 50.48(c) 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 affected by this change. The proposed
license amendment does not adversely affect existing plant safety
margins or the reliability of equipment assumed in the USAR to
mitigate accidents. The proposed change does not adversely impact
systems that respond to safely shut down the plant and maintain the
plant in a safe shutdown condition. In addition, the proposed
license amendment will not result in plant operation in a
configuration outside the design basis for an unacceptable period of
time without implementation of appropriate compensatory measures.
The purpose of the proposed license amendment is to permit CNS
to adopt a new fire protection licensing basis which complies with
the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) and the
guidance in Regulatory Guide 1.205. The NRC considers that NFPA 805
provides an acceptable methodology and performance criteria for
licensees to identify fire protection systems and features that are
an acceptable alternative to the 10 CFR part 50, appendix R required
fire protection features (69 FR 33536; June 16, 2004).
The risk evaluations for plant changes, in part as they relate
to the potential for reducing a safety margin, were measured
quantitatively for acceptability using the delta risk guidance
contained in RG 1.205. Engineering analyses, which may include
engineering evaluations, probabilistic safety assessments, and fire
modeling calculations, have been performed to demonstrate that the
performance-based methods of NFPA 805 do not result in a significant
reduction in the margin of safety.
As such, the proposed changes are evaluated to ensure that risk
and safety margins are kept within acceptable limits. Based on the
above, the implementation of this amendment to transition the FPP at
CNS to one based on NFPA 805, in accordance with 10 CFR 50.48(c),
will not significantly reduce a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: February 28, 2013.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3.6.5, ``Containment Air
Temperature,'' to increase the allowable containment average
temperature from 120 [deg]F to 125 [deg]F. The revised TS Section 3.6.5
would read as follows: ``Containment average air temperature shall be
<=125 [deg]F.''
The licensee supports the proposed change by revising the analyses
for Loss of Coolant Accident (LOCA) and a Main Steam Line break, and
evaluating the containment response by either increase in initial
containment air temperature or increase in the temperature of safety
injection accumulators, which are located in the Ginna containment, and
are expected to be at the same temperature as containment air.
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 to increase the containment average air
temperature limit to 125 [deg]F, from 120 [deg]F, does not alter the
assumed initiators to any analyzed event. The probability of an
accident previously evaluated will not be increased by this proposed
change. This proposed change will not affect radiological dose
consequence analyses. The radiological dose consequence analyses
assume a certain containment atmosphere leak rate based on the
maximum allowable containment leakage rate, which is not affected by
the change in allowable average containment air temperature
resulting in a higher calculated peak containment pressure. The 10
CFR part 50, appendix J containment leak rate testing program will
continue to ensure that containment leakage remains within the
leakage assumed in the offsite dose consequence analyses. The
acceptable leakage corresponds to the peak allowable containment
pressure of 60 psig. The radiological dose consequence analyses
assume a certain source term, which is not affected by the change in
allowable average containment air temperature. All core limitations
set forth in 10 CFR 50.46 continue to be met. The consequences of an
accident previously evaluated will not be increased by this proposed
change.
Therefore, operation of the facility in accordance with the
proposed change to the containment average air temperature limit
will 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 provides for a higher allowable containment
average air
[[Page 70595]]
temperature to that currently in the TS Section 3.6.5. The
calculated peak containment temperature and pressure remain below
the containment design temperature and pressure of 286[emsp14][deg]F
and 60 psig. This change does not involve any alteration in the
plant configuration (no new or different type of equipment will be
installed) or make changes in the methods governing normal plant
operation. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, operation of the facility in accordance with the
proposed change to TS Section 3.6.5 would not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The calculated peak containment pressure and temperature remain
below the containment design pressure and temperature of 60 psig and
286[emsp14][deg]F, respectively. The penalties applied to the BE
LBLOCA [best estimate loss of coolant accident] analysis result in
the limitations set forth in 10 CFR 50.46 continuing to be met.
Since the radiological consequence analyses are based on the maximum
allowable containment leakage rate, which is not being revised, the
change in the calculated peak containment pressure and temperature
and changes in core response do not represent a significant change
in the margin of safety. The longterm impact of the peak containment
temperature following a design basis accident exceeding the EQ
profile by 2[emsp14][deg]F with respect to the current licensing
basis is negligible.
Therefore, operation of the facility in accordance with the
proposed change to increase the allowable containment average air
temperature from 120[emsp14][deg]F to 125[emsp14][deg]F does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor,
Baltimore, MD 21202.
NRC Acting Branch Chief: Robert Beall.
South Carolina Electric and Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of amendment request: October 3, 2013.
Description of amendment request: The proposed amendment would
revise the scheduled completion date of the Cyber Security Plan
Milestone 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Cyber Security Plan
Implementation Schedule. This change does not alter accident
analysis assumptions, add any initiators, or affect the function of
plant systems or the manner in which systems are operated,
maintained, modified, tested, or inspected.
The proposed change is a change to the completion date of
implementation milestone 8 that in itself does not require any plant
modifications which affect the performance capability of the
structures, systems, and components relied upon to mitigate the
consequences of postulated accidents and have no impact on 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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the Cyber Security Plan
Implementation Schedule. This proposed change to modify the
completion date of implementation milestone 8 does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. The proposed
change does not require any plant modifications which affect the
performance capability of the structures, systems and components
relied upon to mitigate the consequences of postulated accidents.
This change also does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed change
revises the Cyber Security Plan Implementation Schedule. Because
there is no change to these established safety margins as result of
this change, the proposed change 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: J. Hagood Hamilton, Jr., South Carolina
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina
29218.
NRC Branch Chief: Robert J. Pascarelli.
South Carolina Electric and Gas Docket Nos. 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: October 2, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station (VCSNS) Units 2 and 3 by departing from the plant-
specific Design Control Document (DCD) Tier 1 (and corresponding
Combined License Appendix C information) and Tier 2 material by making
changes to the Non-Class 1E dc and Uninterruptible Power Supply System
(EDS) and Uninterruptible Power Supply System (IDS) and making changes
to the corresponding Tier 1 information in Appendix C to the Combined
License. The proposed changes would:
(1) Increase EDS total equipment capacity, component ratings, and
protective device sizing to support increased load demand,
(2) Relocate equipment and moving Turbine Building (TB) first bay
EDS Battery Room and Charger Room. The floor elevation increases from
elevation 148'-0'' to elevation 148'-10'' to accommodate associated
equipment cabling with this activity, and
(3) Remove the Class 1E IDS Battery Back-up tie to the Non-Class 1E
EDS Battery. Because, this proposed change requires a departure from
Tier 1 information in the Westinghouse Advanced Passive 1000 DCD, the
licensee also requested an exemption from the requirements of the
Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).
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.
[[Page 70596]]
The design function of the Turbine Building (TB) is to provide
weather protection for the laydown and maintenance of major turbine/
generator components. The TB first bay is a seismic Category II
structure designed to prevent the collapse under a safe shutdown
earthquake (SSE) to protect the adjacent auxiliary building. The
electrical system and air-handling units are designed to provide
electrical power to plant loads and maintain acceptable temperatures
for electrical equipment rooms and work areas. The electrical
equipment continues to be in accordance with the same codes and
standards stated in the Updated Final Safety Analysis Report
(UFSAR). The proposed relocation of equipment, including the
increase in floor elevation by 10 inches to accommodate overhead
equipment cabling, does not impact the TB design function. The TB
first bay continues to meet seismic Category II requirements. Based
on this, the proposed changes would not increase the probability of
an accident previously evaluated.
The proposed changes do not involve any accident initiating
event, thus the probabilities of the accidents previously evaluated
are not affected. The relocation of equipment does not involve any
safety-related structures, systems, or components; the affected
rooms do not represent a radioactive material barrier; and this
activity does not affect the containment of radioactive material.
The radioactive material source terms and release paths used in the
safety analyses are unchanged, thus the radiological releases in the
accident analyses are not affected. Therefore, the consequences of
an accident previously evaluated are not affected.
Therefore, there is no significant increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes would use the same type of electrical
equipment with higher ratings and capacity, change the source of a
battery back-up, and relocate equipment. The electrical equipment
will continue to perform its design functions because the same
electrical codes and standards as stated in the UFSAR continue to be
met. Therefore the proposed changes do not affect equipment failure
probabilities or alter any accident initiator or initiating sequence
of events. The proposed changes in location of equipment and
elevation of the TB first bay floor do not affect the design
function of the TB first bay to protect the adjacent auxiliary
building by meeting seismic Category II structure requirements, or
affect the operation of the relocated equipment, or the ability of
the relocated equipment to meet its design functions. Because the
SSCs and equipment affected by the proposed changes continue to meet
their design functions, the structural codes and standards as stated
in the UFSAR, the proposed changes do not introduce a different type
of accident than those previously considered.
Therefore, this activity 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 current seismic requirements applicable to the seismic
Category II TB first bay structure, including the seismic modeling
and analysis methods, will continue to apply to the TB first bay
floor elevation increase. The proposed changes to relocate equipment
and the increase in the floor elevation will continue to meet the
fire rating requirements and will be in accordance with the same
codes and standards currently identified in the UFSAR. The proposed
changes to the electrical equipment will continue to meet existing
electrical equipment industry standard recommendations identified in
the UFSAR. Because no safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by these proposed changes,
no margin of safety is reduced.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart.
Dated at Rockville, Maryland, this 15th day of November 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-28225 Filed 11-25-13; 8:45 am]
BILLING CODE 7590-01-P