[Federal Register Volume 79, Number 169 (Tuesday, September 2, 2014)]
[Notices]
[Pages 52059-52072]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-20671]
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NUCLEAR REGULATORY COMMISSION
[NRC-2014-0193]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: 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 August 7, 2014 to August 20, 2014. The last
biweekly notice was published on August 19, 2014.
DATES: Comments must be filed by October 2, 2014. A request for a
hearing must be filed by November 3, 2014.
ADDRESSES: You may submit comments 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-2014-0193. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Angela Baxter, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-2976, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2014-0193 when contacting the NRC
about the availability of information for this action. You may obtain
publicly-available information related to this action by any of the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0193.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection 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 (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section.
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-2014-0193 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.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 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)
[[Page 52060]]
create the possibility of a new or different kind of accident from any
accident previously evaluated; or (3) involve a significant reduction
in a margin of safety. The basis for this proposed determination for
each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
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's 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.
B. Electronic Submissions (E-Filing)
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
ten 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.
[[Page 52061]]
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/getting-started.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with 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 NRC'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 public
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 in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress Inc., Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, New Hill, North Carolina
Date of amendment request: June 19, 2014. A publicly-available
version is in ADAMS under Accession No. ML14174A118.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.2, ``Engineered Safety Features
Actuation System Instrumentation,'' Table 3.3-4, ``Engineered Safety
Features Actuation System Instrumentation Trip Setpoints.''
Specifically, the instrument trip setpoint and associated allowable
value are being revised to ensure that the trip of the safety-related
alternating current bus will occur at a voltage at or above the minimum
voltage necessary to operate the applicable safety-related loads.
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 TS Table 3.3-4 Functional Unit
9.a, Loss-of-Offsite Power 6.9 kV Emergency Bus Undervoltage--
Primary, instrumentation trip setpoint and
[[Page 52062]]
allowable value. The Loss-of-Offsite Power, 6.9 kV Emergency Bus
Undervoltage--Primary instrumentation is not an initiator to any
accident previously evaluated. As such, the probability of an
accident previously evaluated is not increased. The Loss-of-Offsite
Power, 6.9 kV Emergency Bus Undervoltage--Primary instrumentation
revised values continue to provide reasonable assurance that the
Functional Unit 9.a will continue to perform its intended safety
functions. As a result, the proposed change will not increase the
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 TS Table 3.3-4 Functional Unit
9.a, Loss-of-Offsite Power 6.9 kV Emergency Bus Undervoltage--
Primary, instrumentation trip setpoint and allowable value. No new
operational conditions beyond those currently allowed are
introduced. This change is consistent with the safety analyses
assumptions and current plant operating practices. This simply
corrects the setpoint consistent with the accident analyses and
therefore cannot create the possibility of a new or different kind
of accident from any previously evaluated accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the TS Table 3.3-4 Functional Unit
9.a, Loss-of-Offsite Power 6.9 kV Emergency Bus Undervoltage--
Primary, instrumentation trip setpoint and allowable value. Function
9.a protects the emergency power system against loss of voltage.
This change is consistent with the safety analyses assumptions and
current plant operating practices. No new operational conditions
beyond those currently allowed are created by these changes.
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, Deputy General Counsel,
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A,
Charlotte, NC 28202.
NRC Acting Branch Chief: Lisa M. Regner.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of amendment request: June 11, 2014. A publicly-available
version is in ADAMS under Accession No. ML14162A079.
Description of amendment request: The proposed amendment would
modify technical specification (TS) requirements to adopt the changes
described in TS Task Force (TSTF)-426, Revision 5, ``Revise or Add
Actions to Preclude Entry into LCO [limiting condition for operation]
3.0.3--RITSTF [Risk-Informed TSTF] Initiatives 6b & 6c'' (ADAMS
Accession No. ML113260461).
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 provides a short Completion Time to restore
an inoperable system for conditions under which the existing
Technical Specifications require a plant shutdown to begin within 1
hour in accordance with LCO 3.0.3. Entering into Technical
Specification Actions is not an initiator of any accident previously
evaluated. As a result, the probability of an accident previously
evaluated is not significantly increased. The consequences of any
accident previously evaluated that may occur during the proposed
Completion Times are no different from the consequences of the same
accident during the existing 1 hour allowance. As a result, the
consequences of any accident previously evaluated are not
significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different type of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change increases the time the plant may operate
without the ability to perform an assumed safety function. The
analysis in WCAP-16125-NP-A, ``Justification for Risk-Informed
Modifications to Selected Technical Specifications for Conditions
Leading to Exigent Plant Shutdown,'' Revision 2, August 2010,
demonstrated that there is an acceptably small increase in risk due
to a limited period of continued operation in these conditions and
that the risk is balanced by avoiding the risks associated with a
plant shutdown. As a result, the change to the margin of safety
provided by requiring a plant shutdown within 1 hour is not
significant.
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 Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: David L. Pelton.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of amendment request: July 11, 2014. A publicly-available
version is in ADAMS under Accession No. ML14192B143.
Description of amendment request: The proposed amendment would
incorporate several miscellaneous administrative changes to the
Facility Operating License and the Technical Specifications. For
example, the amendment would delete historical items that are no longer
applicable, correct errors, and remove references that are no longer
valid.
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.
No physical changes to the facility will occur as a result of
this proposed amendment. The proposed changes will not alter the
physical design or operational procedures associated with any plant
structure, system, or component. The proposed changes are
administrative in nature and have no effect on plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
[[Page 52063]]
Response: No.
The proposed changes are administrative in nature. The proposed
changes do not alter the physical design, safety limits, or safety
analysis assumptions associated with the operation of the plant.
Accordingly, the changes do not introduce any new accident
initiators, nor do they reduce or adversely affect the capabilities
of any plant structure, system, or component to perform their safety
function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes conform to NRC regulatory guidance
regarding the content of plant Technical Specifications. The
proposed changes are administrative in nature. The proposed changes
do not alter the physical design, safety limits, or safety analysis
assumptions associated with the operation of the plant.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: J. Bradley Fewell, Vice President and Deputy
General Counsel, Exelon Generation Company, LLC, 200 Exelon Way,
Kennett Square, PA 19348.
Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: July 10, 2014. A publicly-
available version is in ADAMS under Accession No. ML14191B190.
Description of amendment request: The proposed amendment would
revise and add Technical Specification (TS) surveillance requirements
to address the concerns discussed in NRC Generic Letter 2008-01,
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013
(ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of
Availability for TSTF-523, Revision 2, for plant-specific adoption
using the Consolidated Line Item Improvement Process, in the Federal
Register on January 15, 2014 (79 FR 2700).
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 or adds Surveillance Requirements
(SRs) that require verification that the Emergency Core Cooling
Systems, the Suppression Pool Cooling System, the Suppression Pool
Spray System, the Drywell Spray System, the Shutdown Cooling System,
and the Reactor Core Isolation Cooling System are not rendered
inoperable due to accumulated gas and to provide allowances which
permit performance of the revised verification. Gas accumulation in
the subject systems is not an initiator of any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The proposed SRs ensure
that the subject systems continue to be capable of performing their
assumed safety function and are not rendered inoperable due to gas
accumulation. Thus, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises or adds SRs that require
verification that the Emergency Core Cooling Systems, the
Suppression Pool Cooling System, the Suppression Pool Spray System,
the Drywell Spray System, the Shutdown Cooling System, and the
Reactor Core Isolation Cooling System are not rendered inoperable
due to accumulated gas and to provide allowances which permit
performance of the revised verification. The proposed 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 operation. In addition, the proposed
change does not impose any new or different requirements that could
initiate an accident. The proposed change does not alter assumptions
made in the safety analysis and is consistent with the safety
analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises or adds SRs that require
verification that the Emergency Core Cooling Systems, the
Suppression Pool Cooling System, the Suppression Pool Spray System,
the Drywell Spray System, the Shutdown Cooling System, and the
Reactor Core Isolation Cooling System are not rendered inoperable
due to accumulated gas and to provide allowances which permit
performance of the revised verification. The proposed change adds
new requirements to manage gas accumulation in order to ensure the
subject systems are capable of performing their assumed safety
functions. The proposed SRs are more comprehensive than the current
SRs and will ensure that the assumptions of the safety analysis are
protected. The proposed change does not adversely affect any current
plant safety margins or the reliability of the equipment assumed in
the safety analysis. Therefore, there are no changes being made to
any safety analysis assumptions, safety limits or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed change.
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. Bradley Fewell, Esquire, Vice President
and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon
Way, Kennett Square, PA 19348.
Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: July 10, 2014. A publicly-available
version is in ADAMS under Accession No. ML14191A059.
Description of amendment request: The proposed amendment would
revise and add Technical Specification (TS) Surveillance Requirements
to address the concerns discussed in NRC Generic Letter 2008-01,
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013
(ADAMS Accession
[[Page 52064]]
No. ML13053A075). The NRC staff issued a Notice of Availability for
TSTF-523, Revision 2, for plant-specific adoption using the
Consolidated Line Item Improvement Process, in the Federal Register on
January 15, 2014 (79 FR 2700).
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 adds Surveillance Requirements (SRs) that
require verification that the Emergency Core Cooling System (ECCS),
the Decay Heat Removal (DHR) System, and the Reactor Building Spray
(RB Spray) System are not rendered inoperable due to accumulated gas
and to provide allowances which permit performance of the revised
verification. Gas accumulation in the subject systems is not an
initiator of any accident previously evaluated. As a result, the
probability of any accident previously evaluated is not
significantly increased. The proposed SRs ensure that the subject
systems continue to be capable of performing their assumed safety
function and are not rendered inoperable due to gas accumulation.
Thus, the consequences of any accident previously evaluated are not
significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change adds SRs that require verification that the
ECCS, the DHR, and the RB Spray System are not rendered inoperable
due to accumulated gas and to provide allowances which permit
performance of the revised verification. The proposed 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 operation. In addition, the proposed
change does not impose any new or different requirements that could
initiate an accident. The proposed change does not alter assumptions
made in the safety analysis and is consistent with the safety
analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adds SRs that require verification that the
ECCS, the DHR, and the RB Spray System are not rendered inoperable
due to accumulated gas and to provide allowances which permit
performance of the revised verification. The proposed change adds
new requirements to manage gas accumulation in order to ensure that
the subject systems are capable of performing their assumed safety
functions. The proposed SRs are more comprehensive than the current
SRs and will ensure that the assumptions of the safety analysis are
protected. The proposed change does not adversely affect any current
plant safety margins or the reliability of the equipment assumed in
the safety analysis. Therefore, there are no changes being made to
any safety analysis assumptions, safety limits, or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed change.
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. Bradley Fewell, Vice President and Deputy
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: July 10, 2014. A publicly-available
version is in ADAMS under Accession No. ML14191B180.
Description of amendment request: The proposed amendment would
revise and add Technical Specification (TS) surveillance requirements
to address the concerns discussed in NRC Generic Letter 2008-01,
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013
(ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of
Availability for TSTF-523, Revision 2, for plant-specific adoption
using the Consolidated Line Item Improvement Process, in the Federal
Register on January 15, 2014 (79 FR 2700).
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 or adds Surveillance Requirements
(SRs) that require verification that the Emergency Core Cooling
System (ECCS), the Residual Heat Removal (RHR) System, the Shutdown
Cooling (SDC) System, the Containment Spray (CS) System, and the
Reactor Core Isolation Cooling (RCIC) System are not rendered
inoperable due to accumulated gas and to provide allowances which
permit performance of the revised verification. Gas accumulation in
the subject systems is not an initiator of any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The proposed SRs ensure
that the subject systems continue to be capable of performing their
assumed safety function and are not rendered inoperable due to gas
accumulation. Thus, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, the RHR, the SDC, the CS, and the RCIC
Systems are not rendered inoperable due to accumulated gas and to
provide allowances which permit performance of the revised
verification. The proposed 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
operation. In addition, the proposed change does not impose any new
or different requirements that could initiate an accident. The
proposed change does not alter assumptions made in the safety
analysis and is consistent with the safety analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, the RHR, the SDC, the CS, and the RCIC
Systems are not rendered inoperable due to accumulated gas and to
provide allowances which permit performance of the revised
verification. The proposed change revises or adds new requirements
to manage gas accumulation in order to ensure the subject
[[Page 52065]]
systems are capable of performing their assumed safety functions.
The proposed SRs are more comprehensive than the current SRs and
will ensure that the assumptions of the safety analysis are
protected. The proposed change does not adversely affect any current
plant safety margins or the reliability of the equipment assumed in
the safety analysis. Therefore, there are no changes being made to
any safety analysis assumptions, safety limits or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed change.
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. Bradley Fewell, Vice President and Deputy
General Counsel, Exelon Generation Company, LLC, 200 Exelon Way,
Kennett Square, PA 19348.
Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2,
Ogle County, Illinois
Date of amendment request: April 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14120A039.
Description of amendment request: The proposed amendment would add
new ``low degraded voltage relays'' and timers, with appropriate
settings, on each engineered safety feature electrical bus. The
technical specifications and surveillance requirements would be changed
to add appropriate operational and testing requirements for the new
relays and timers.
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, with NRC staff revisions provided in [brackets], which
is presented below:
EGC [Exelon Generation Company, LLC] has evaluated the proposed
change for Braidwood Station and Byron Station, using the criteria
in 10 CFR 50.92, and has determined that the proposed change does
not involve a significant hazards consideration. The following
information is provided to support a finding of no significant
hazards consideration.
Criteria
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 add new ``low degraded voltage relays''
(LDVRs) and associated CHANNEL CALIBRATION surveillance test
provides a third level of undervoltage protection for the Engineered
Safeguards Features (ESF) electrical buses. These new relays will
further ensure that the normally operating safety-related motors/
equipment, which are powered from the ESF buses, are appropriately
isolated from the normal off-site power source and will not be
damaged in the event of sustained degraded bus voltage. The addition
of the LDVRs will continue to allow the existing undervoltage
protection circuitry to function as originally designed; i.e., the
first-level ``loss of voltage'' protection and the second-level
``degraded voltage'' protection will remain in place and be
unaffected by this change. The proposed change does not affect the
probability of any accident resulting in a loss of voltage or
degraded voltage condition on the ESF electrical buses; and will
positively impact the consequences of accidents previously evaluated
as this change further ensures continued operation of safety-related
equipment throughout the accident scenarios.
Specific analysis was performed and determined that the proposed
LDVRs, with the specified allowable values and time delay, will
ensure that the 4.16 kV ESF buses will be isolated from the normal
off-site power source, at the appropriate voltage level, under
nonaccident sustained degraded voltage conditions. The normally
operating safety related motors will be subsequently sequenced back
on to the 4.16 kV ESF buses powered by the EDGs [Emergency Diesel
Generators]; and therefore, will not be damaged in the event of
sustained degraded bus voltage during the time delay period prior to
initiation of the first level loss of voltage trip function.
Therefore, these safety-related loads will be available to
perform their design basis function should a loss-of-coolant
accident (LOCA) occur concurrent with a loss-of-offsite power (LOOP)
following the degraded voltage condition. The loading sequence
(i.e., timing) of safety-related equipment back onto the ESF bus,
powered by the EDG, is not affected by the addition of the new
LDVRs.
The addition of new LDVRs will have no impact on accident
initiators or precursors; does not alter the accident analysis
assumptions or the manner in which the plant is operated or
maintained; and does not affect the probability of operator error.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change involves the addition of new ``low degraded
voltage relays'' (LDVRs); i.e., a third level of undervoltage
protection for the ESF electrical buses, and adds an associated
CHANNEL CALIBRATION surveillance test. This change helps ensure that
the assumptions in the previously evaluated accidents, which may
involve a degraded voltage condition, continue to be valid.
The proposed changes do not result in the creation of any new
accident precursors; do not result in changes to any existing
accident scenarios; and do not introduce any operational changes or
mechanisms that would create the possibility of a new or different
kind of accident. A specific failure mode and effects review was
completed for the new LDVRs, considering their potential failure,
and concluded that the addition of these relays would not affect the
existing ``loss of voltage'' and ``degraded voltage'' protection
schemes; would not affect the number of occurrences of degraded
voltage conditions that would cause the actuation of the existing
Loss of Voltage Relays (LVRs), Degraded Voltage Relays (DVRs) or new
LVDRs; would not affect the failure rate of the existing protection
relays; and would not impact the assumptions in any existing
accident scenario.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The current ``loss of voltage'' and ``degraded voltage''
protection circuitry is designed to appropriately isolate the
normally operating safety-related motors/equipment, which are
powered from the ESF buses, from the normal off-site power source
such that the subject equipment will not be damaged in the event of
sustained degraded bus voltage. The loss of voltage relays (LVRs)
isolate the ESF buses at a TS [technical specifications] voltage
value of approximately 66% of the nominal bus value after a short
time delay (i.e., 1.9 seconds); while the degraded voltage relays
(DVRs) isolate the ESF buses at a TS voltage value of 94.5% for
Braidwood (91.2% for Byron Station) of the nominal bus voltage after
a longer time delay of up to 5 minutes and 40 seconds (if no safety
injection signal is present). After the ESF buses are isolated from
the offsite power supply, the normally operating safety related
motors will be sequenced back on to the 4.16 kV EFS bus powered by
the EDG; and continue to perform their design basis function to
mitigate the consequences of an accident, with a specified margin of
safety.
A concern exists that ESF motors/equipment may be damaged when
operating and/or starting safety-related equipment when bus voltage
drops to just above the loss of voltage relay setpoint for the
duration of the 5 minutes and 40 second time delay. The new LDVRs
are being added to resolve this concern. Analysis has been performed
that shows the ESF equipment will not be damaged at 75% of bus
voltage; therefore, the LDVR setpoint will be set at 75% of nominal
ESF bus voltage. With the addition of this new third level of
undervoltage protection,
[[Page 52066]]
the capability of the ESF equipment will be assured; and thus the
equipment will continue to perform its design basis function to
mitigate the consequences of the previously analyzed accidents; and
maintain the existing margin to safety currently assumed in the
accident analyses.
An EDG start due to a safety injection signal (i.e., Loss of
Coolant Accident) and the subsequent sequencing of ESF loads back on
to the ESF buses, powered by the EDG, is not adversely affected by
this change. If an actual loss of voltage condition occurs on the
ESF buses, the loss of voltage time delays will continue to isolate
the 4.16 kV ESF distribution system from the offsite power source
prior to the EDG assuming the ESF loads.
The ESF loads will sequence back on to the bus in a specified
order and time interval; again ensuring that the existing accident
analysis assumptions remain valid and the existing margin to safety
is unaffected.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, EGC concludes that the proposed amendments
do not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: June 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14177A503.
Description of amendment request: The proposed amendment would
revise and add Technical Specification (TS) Surveillance Requirements
(SRs) to address the concerns discussed in NRC Generic Letter 2008-01,
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013
(ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of
Availability for TSTF-523, Revision 2, for plant-specific adoption
using the Consolidated Line Item Improvement Process, in the Federal
Register on January 15, 2014 (79 FR 2700).
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, with NRC staff revisions provided in [brackets], 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 or adds SRs that require
verification that the Emergency Core Cooling Systems (ECCS),
Residual Heat Removal (RHR) System, and Containment Spray (CS)
System are not rendered inoperable due to accumulated gas and to
provide allowances which permit performance of the revised
verification. Gas accumulation in the subject systems is not an
initiator of any accident previously evaluated. As a result, the
probability of any accident previously evaluated is not
significantly increased. The proposed SRs ensure that the subject
systems continue to be capable to perform their assumed safety
function and are not rendered inoperable due to gas accumulation.
Thus, the consequences of any accident previously evaluated are not
significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, RHR System, and CS System are not
rendered inoperable due to accumulated gas and to provide allowances
which permit performance of the revised verification. The proposed
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 operation. In addition, the
proposed change does not impose any new or different requirements
that could initiate an accident. The proposed change does not alter
assumptions made in the safety analysis and is consistent with the
safety analysis assumptions[.]
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, RHR System, and CS System are not
rendered inoperable due to accumulated gas and to provide allowances
which permit performance of the revised verification. The proposed
change adds new requirements to manage gas accumulation in order to
ensure that the subject systems are capable of performing their
assumed safety functions. The proposed SRs are more comprehensive
than the current SRs and will ensure that the assumptions of the
safety analysis are protected. The proposed change does not
adversely affect any current plant safety margins or the reliability
of the equipment assumed in the safety analysis. Therefore, there
are no changes being made to any safety analysis assumptions, safety
limits, or limiting safety system settings that would adversely
affect plant safety as a result of the proposed change.
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: James Petro, Managing Attorney, Florida
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
Acting NRC Branch Chief: Robert G. Schaaf.
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: May 20, 2014, as supplemented by letter
dated June 3, 2014. Publicly-available versions are in ADAMS under
Accession Nos. ML14140A637 and ML14155A257, respectively.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-93 and NPF-94 for VCSNS Units 2 and 3 by
departing from the plant-specific Design Control Document (DCD) Tier
1(and corresponding Combined License Appendix C information) material
by making various nontechnical changes to correct editorial and
consistency errors in Tier 1. This is being done to promote consistency
within the Updated Final Safety Analysis Report (UFSAR).
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
[[Page 52067]]
licensee has provided its analysis of the issue of no significant
hazards consideration, with NRC staff revisions provided in [brackets],
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 editorial and consistency plant-specific Tier 1 and
corresponding COL [combined operating license] Appendix C update
does not involve a technical change, e.g., there is no design
parameter or requirement, calculation, analysis, function or
qualification change. No structure, system, or component (SSC)
design or function would be affected. No design or safety analysis
would be affected. The proposed changes do not affect any accident
initiating event or component failure, thus the probabilities of the
accidents previously evaluated are not affected. No function used to
mitigate a radioactive material release and no radioactive material
release source term is involved, thus the radiological releases in
the accident analyses are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed editorial and consistency plant-specific Tier 1 and
corresponding COL Appendix C update would not affect the design or
function of any SSC, but will instead provide consistency between
the SSC designs and functions currently presented in the UFSAR and
the Tier 1 information. The proposed changes would not introduce a
new failure mode, fault or sequence of events that could result in a
radioactive material release.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed editorial and consistency plant-specific Tier 1 and
corresponding COL Appendix C update is considered non-technical for
reasons discussed above, thus would not affect any design parameter,
function or analysis. There would be no change to an existing design
basis, design function, regulatory criterion, or analysis. No safety
analysis or design basis acceptance limit/criterion is involved.
Therefore, the proposed amendment does not reduce 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: Kathryn M. Sutton, Morgan, Lewis & Bockius
LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2 (VEGP), Burke
County, Georgia
Date of amendment request: August 31, 2012, as supplemented
September 13, 2013, May 2, July 22, and August 11, 2014. Publicly-
available versions are in ADAMS under Accession Nos. ML12248A035,
ML13256A306, ML14122A364, ML14203A252 and, ML14223A616, respectively.
Description of amendment request: The proposed amendments would
revise the licensing basis for the VEGP by adding license conditions
that would allow for the voluntary implementation of 10 CFR 50.69,
``Risk-informed categorization and treatment of structures, systems,
and components for nuclear power reactors.'' As indicated in Sec.
50.69, a licensee may voluntarily comply with Sec. 50.69 as an
alternative to compliance with the following requirements for certain
SSCs: (i) 10 CFR part 21, (ii) a portion of Sec. 50.46, (iii) Sec.
50.49, (v) certain requirements of Sec. 50.55a, (vi) Sec. 50.65,
(vii) Sec. 50. 72, (viii) Sec. 50.73,[middot](ix) Appendix B to Part
50, (x) certain containment leakage testing requirements, and (xi)
certain requirements of Appendix A to part 100.
Basis for proposed no significant hazards consideration
determination: The licensee responded in its letter dated August 11,
2014, to the NRC staff's request for additional information regarding
the licensee's no significant hazards consideration determination,
which is required by 10 CFR 50.91(a). Portions of the licensee's
response regarding each of the no significant hazards consideration
standards, with NRC staff revisions provided in [brackets], are
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 Vogtle Electric Generating Plant (VEGP) in
accordance with the proposed amendment does not result in a
significant increase in the probability or consequences of accidents
previously evaluated. The Updated Final Safety Analysis Report
(UFSAR) documents the analysis of design basis accidents at VEGP.
The proposed amendment does not affect accident initiators, nor does
it alter design assumptions, conditions, or configurations of the
facility that would increase the probability of accidents previously
evaluated, 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, nor do they affect assumed failure modes for accidents
described and evaluated in the UFSAR. The proposed changes do not
affect the way in which required systems perform their functions as
required by the accident analysis. Structures, systems, and
components required to safely shut down the reactor and maintain it
in a safe shutdown condition will remain capable of performing their
design functions.
Furthermore, the source term and radiological release
assumptions of previously evaluated events are not affected by the
alternative treatments permitted under 10 CFR 50.69; containment
isolation devices assumed to function under accident conditions will
not have their reliability adversely affected by the proposed
amendment. Consequently, operating under the proposed amendment will
not result in a significant increase in the radiological dose
consequences assumed for previously analyzed events.
Section 50.69 defines the terminology ``safety significant
function'' as functions whose loss or degradation could have a
significant adverse effect on defense-in-depth, safety margins, or
risk. For SSCs determined to be safety significant, 50.69 maintains
the current regulatory requirements. These current requirements are
adequate for addressing design basis performance of these SSCs.
The purpose of this amendment is to permit VEGP to adopt a new
risk-informed licensing basis for categorization and treatment of
structures, systems and components. The proposed VEGP Units 1 and 2
OL [operating license] LCs [license conditions] will allow for the
voluntary implementation of 10 CFR 50.69. The SNC [Southern Nuclear
Operating Company] risk-informed categorization process has been
documented per the requirements of 10 CFR 50.69(b)(2) and meets the
requirements of 10 CFR 50.69(c). A probabilistic approach to
regulation enhances and extends the traditional deterministic
approach by allowing consideration of a broader set of potential
challenges to safety and providing a logical means for prioritizing
these challenges based on safety significance. The SNC risk-informed
categorization process will be used to modify the scope of SSCs
subject to special treatment requirements. Alternative treatments
permitted per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) can then be
applied consistent with the categorization of the SSCs. The process
provides reasonable confidence that, for SSCs categorized as RISC-3,
sufficient safety margins are maintained and that any potential
increases in CDF [core damage frequency] and LERF [large early
release frequency] resulting from changes in treatment are small per
10 CFR 50.69(c)(1)(iv). The proposed OL LCs do not result in or
require any physical or operational changes to VEGP SSCs, including
SSCs intended for the prevention or
[[Page 52068]]
mitigation of accidents. Implementation of 10 CFR 50.69 in
compliance with 10 CFR 50.69 requirements ensures that RISC-1 and
RISC-3 SSCs remain capable of performing their design basis
functions, including safety-related functions, under design basis
conditions. In addition, the process ensures that RISC-2 SSCs are
capable of performing their safety significant functions.
Based on the above, implementation of this amendment to
implement 10 CFR 50.69 risk informed categorization and treatment of
structures, systems, and components does not involve 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.
Therefore, consequences of any accident previously evaluated are
not significantly increased with the implementation of this License
Amendment.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Operation of VEGP in accordance with the proposed amendment does
not create the possibility of a new or different kind of accident
from any accident previously evaluated. The proposed amendment does
not impact any scenario or previously analyzed accident with offsite
dose consequences included in the evaluation of design basis
accidents (DBA) documented in the FSAR [final safety analysis
report]. The proposed change does not alter the requirements or
functions for systems required during accident conditions, nor does
it alter the required mitigation systems as assumed in the licensing
basis analyses and/or DBA radiological consequences evaluations.
Implementation of the 50.69 categorization will not result in new or
different accidents.
The proposed amendment does not adversely affect accident
initiators nor alter design assumptions, or conditions of the
facility. The proposed amendment does not introduce new or different
accident initiators; neither does it introduce new modes of
operation. The proposed amendment does not adversely affect the
ability of SSCs to perform their design function. SSCs required to
safely shutdown the reactor and maintain it in a safe shutdown
condition remain capable of performing their design function.
Section 50.69 represents an alternative set of requirements
whereby a licensee may voluntarily undertake categorization of its
SSCs consistent with the requirements in 50.69(c), remove the
special treatment requirements listed in 50.69(b) for SSCs that are
determined to be of low safety significance, and implement
alternative treatment requirements in 50.69(d). The regulatory
requirements not removed continue to apply. These requirements are
adequate for addressing design basis performance of these SSCs. This
license amendment continues to maintain the principles that the net
increase in plant risk is small, defense-in-depth is maintained, and
safety margins are maintained.
The proposed VEGP Units 1 and 2 OL LCs will allow for the
voluntary implementation of 10 CFR 50.69. The SNC risk-informed
categorization process has been documented per the requirements of
10 CFR 50.69(b)(2) and meets the requirements of 10 CFR 50.69(c).
The SNC risk-informed categorization process will be used to modify
the scope of SSCs subject to special treatment requirements.
Alternative treatments permitted per 10 CFR 50.69(b)(1) and 10 CFR
50.69(d)(2) can then be applied consistent with the categorization
of the SSCs. The process provides reasonable confidence that, for
SSCs categorized as RISC-3, sufficient safety margins are maintained
and that any potential increases in CDF and LERF resulting from
changes in treatment are small per 10 CFR 50.69(c)(1)(iv). The
proposed OL LCs do not result in or require any physical or
operational changes to VEGP SSCs, including SSCs intended for the
prevention or mitigation of accidents. Implementation of 10 CFR
50.69 in compliance with 10 CFR 50.69 requirements ensures that
RISC-1 and RISC-3 SSCs remain capable of performing their design
basis functions, including safety-related functions, under design
basis conditions. In addition, the process ensures that RISC-2 SSCs
are capable of performing their safety significant functions.
Therefore, even though there was not an individual evaluation done
of every UFSAR accident with potential off-site dose consequences,
it can be concluded that the SSCs, assumed to mitigate the
consequences of any and all previously evaluated events, will not be
adversely affected by the alternative treatments allowed under 10
CFR 50.69. Consequently, the dose consequences of previously
analyzed events will not significantly increase as a result of the
alternative treatment of SSCs. Additionally, implementation of 10
CFR 50.69 will not create new failure mechanisms that initiate new
accidents because the process does not result in or require any
physical or operational changes for VEGP SSCs nor does it alter the
functions or functional requirements of those SSCs.
Based on this, implementation of the proposed amendment would
not create the possibility of a new or different kind of accident
from any kind of accident previously evaluated. No new accident
scenarios, transient precursors, failure mechanisms, or limiting
single failures will be introduced as a result of this amendment.
There will be no adverse effect or challenges imposed on required
systems as a result of this amendment.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Operation of VEGP in accordance with the proposed amendment does
not involve a significant reduction in the margin of safety.
Implementation of a new risk informed categorization and treatment
of structures, systems, and components licensing basis that complies
with the requirements of 10 CFR 50.69 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
amendment does not adversely affect existing plant safety margins or
the reliability of equipment assumed in the UFSAR to mitigate
accidents. The proposed change does not adversely affect the ability
of SSCs to perform their design function. The 10 CFR 50.69 process
provides reasonable confidence that SSCs categorized as RISC-1,
RISC-2, and RISC-3 maintain sufficient safety margins. The proposed
amendment does not adversely impact systems required to safely
shutdown the plant and maintain it in a safe condition.
The proposed VEGP Units 1 and 2 OL LCs will allow for the
voluntary implementation of 10 CFR 50.69. The SNC risk-informed
categorization process has been documented per the requirements of
10 CFR 50.69(b)(2) and meets the requirements of 10 CFR 50.69(c).
The SNC risk-informed categorization process will be used to modify
the scope of SSCs subject to special treatment requirements.
Alternative treatments permitted per 10 CFR 50.69(b)(1) and 10 CFR
50.69(d)(2) can then be applied consistent with the categorization
of the SSCs. Although there were no calculations or evaluations
performed for the express purpose of demonstrating that the
implementation of 10 CFR 50.69 will not result in a significant
reduction in the margin of safety, the process provides reasonable
confidence that, for SSCs categorized as RISC-3, sufficient safety
margins are maintained and that any potential increases in CDF and
LERF resulting from changes in treatment are small per 10 CFR
50.69(c)(1)(iv). The only requirements that are relaxed for SSCs,
consistent with their categorization, are those related to
treatment. The safety margins associated with SSCs design basis
functions and design technical requirements remain unchanged.
Additionally, it is required that there be reasonable confidence
that any potential increases in CDF and LERF be small from assumed
changes in reliability resulting from the treatment changes
permitted by 10 CFR 50.69. As a result individual SSCs continue to
be capable of performing their design basis functions. It is
concluded that sufficient safety margins are preserved.
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: Leigh D. Perry, SVP & General Counsel,
Southern Nuclear Operating Company, 40 Inverness Center Parkway,
Birmingham, AL 35242.
NRC Branch Chief: Robert Pascarelli.
[[Page 52069]]
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2 (HNP), Appling County, Georgia
Date of amendment request: August 15, 2014. A publicly-available
version is in ADAMS under Accession No. ML14227A921.
Description of amendment request: The proposed amendments would
modify Technical Specification (TS) 3.8.7 to add two new safety-related
instrument buses to the HNP electrical distribution system. Certain
instruments will be re-located from existing safety-related electrical
instrument buses to these new ``critical instrumentation buses.'' The
existing instrument bus is listed in TS 3.8.7 of the HNP, Units 1 and
2, TSs and, since some of the instruments powered from this bus will be
moved to the critical instrumentation bus, the new bus will be added to
the list of the existing electrical buses in TS 3.8.7.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided an analysis of the issue of no significant hazards
consideration, with NRC staff revisions provided in [brackets], as
presented below:
Southern Nuclear Operating Company has evaluated whether or not
a significant hazards consideration is involved with the proposed
amendment by focusing on the three standards set forth in 10 CFR
50.92, ``Issuance of Amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously
identified?
Response: No.
These new critical instrumentation buses and their inverters are
not intended for the prevention of any previously analyzed transient
or accident. They are intended to provide power to instruments which
may be necessary to aid the operator in the mitigation of a beyond
design basis external event. The new critical instrumentation buses
perform the same function as existing instrumentation buses except
they will have the added capability of obtaining primary power from
DC [direct current] through their inverters connected to the station
service DC power supplies.
The new equipment (inverters and critical instrumentation bus)
will be installed as safety related, seismically and environmentally
qualified equipment, with the primary power coming from the safety
related DC station service buses, and alternate power available from
the safety related AC [alternating current] essential cabinets.
Therefore, the instruments being moved to the critical
instrumentation bus will have a highly reliable source of power.
Consequently, should the operator require the use of one of these
instruments to aid in mitigating the consequences of a previously
analyzed design basis event, it is highly likely that they will be
available to him/her. It is therefore unlikely that the consequences
of a previously evaluated accident would increase due to an
inability to monitor a key containment parameter.
The TSs are being revised to add these instrument buses to the
LCO [limiting condition for operation] requirements for the
electrical distribution buses. No other TS LCOs are changing, no
Surveillance Requirements are changing, and no instrument setpoints
are changing. In fact, this TS change does not reduce any
requirements. All of the components required to be Operable by the
TSs before this revision request, will be required to be Operable
following this change, as well as the new critical instrumentation
bus. The TS requirements will therefore remain the same for the
instruments being powered from the new critical instrumentation bus
as well as for the instruments remaining on the AC instrument buses.
In other words, the power supplies for these instruments will still
be included in the TS as LCO requirements, as they were before the
design change to add the critical instrumentation buses. The TS
requirements will therefore continue to ensure that these indicators
remain Operable during design basis events.
For the above reasons, revising the TS to include the new
critical instrument buses in the electrical bus distribution
Limiting Condition for Operation does not increase the probability,
or consequences, of a previously analyzed event.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
TS LCO 3.8.7 is being changed to add the new critical
instrumentation bus. No new modes of operation or new failure modes
result from the actual TS change to any system intended for the
prevention of accidents.
The design function of the instruments being moved from the
existing instrument buses to the critical instrumentation buses will
not change. Also, the operation of these instruments during any type
of event is not changing. Only their power supply is being changed
and thus no new modes of operation are created for these
instruments. It is true that new components are being introduced,
i.e., the inverters and instrumentation buses, thus introducing a
potential failure that would not be present before the modification.
However, their failure cannot cause a new or different type of
accident. Furthermore the addition of these instruments will not
affect any other system intended for the prevention of accidents.
The design change does not impact the existing essential
cabinets or instrument buses, except to remove some loads from the
instrument bus. Consequently, the design function, operation,
maintenance, and testing of these existing power supplies will not
change.
Finally, the new inverters and the critical instrumentation
buses are not potential accident initiators; they are not intended
to prevent an accident in that they do not serve as a barrier to the
release of radiation either from the direct fission product
boundary, or from the containment. Rather, they are intended to
power instruments which serve the operators in their attempt to
mitigate the consequences of accidents. Therefore, failure of these
power supplies, or failure of any instrument being powered from
them, cannot create an accident.
For the above reasons, the proposed amendment will not create
the possibility of a new or different type of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The new critical instrumentation buses being referenced in the
TS will power several instruments currently being powered by the
safety related instrument bus. The new inverters and critical
instrumentation buses will also be safety related, as will their
primary power source, the DC station service buses. Additionally,
the inverters are alternately powered from the safety related
essential cabinets. Therefore, because of the reliability and
diversity of power supplies, the margin of safety of a loss of power
event to the relocated instruments is not significantly reduced.
Loading calculations confirm that adequate design margin still
exists for the DC station service buses with respect to their
loading for design basis events, even with the additional loads of
the added instruments.
Additionally, area heat load calculations were performed for the
130 foot elevation of the Units 1 and 2 Control Buildings which
account for the new inverters, instrumentation bus and supporting
components. These calculations concluded that there are no adverse
effects on the [Final Safety Analysis Report] FSAR design functions.
Adding the critical instrumentation buses to the TS ensures that
the new power supplies to the safety related instruments have the
same TS requirements as their previous power supply. Therefore, no
TS requirements have been eliminated or reduced.
For the above reasons, the margin of safety is not significantly
reduced.
On the basis of the evaluation above provided by the licensee, the
NRC staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Leigh D. Perry, SVP & General Counsel of
Operations and Nuclear, Southern Nuclear Operating Company, Inc., 40
Inverness Center Parkway, Birmingham, AL 35242.
NRC Branch Chief: Robert Pascarelli.
[[Page 52070]]
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, (NAPS) Louisa County, Virginia
Date of amendment request: June 30, 2014. A publicly-available
version is in ADAMS under Accession No. ML14183B318.
Description of amendment request: The proposed license amendment
requests the changes to the Technical Specification (TS) TS 5.5.15,
``Containment Leakage Rate Testing Program,'' by replacing the
reference to Regulatory Guide (RG) 1.163, ``Performance-Based
Containment Leak-Test Program,'' with a reference to Nuclear Energy
Institute (NEI) topical report NEI 94-01, Revision 3-A, ``Industry
Guideline for Implementing Performance-Based Option of 10 CFR Part 50,
Appendix J,'' as the implementation document used to develop the North
Anna performance-based leakage testing program in accordance with
Option B of 10 CFR Part 50, Appendix J.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The proposed amendment involves changes to the NAPS Containment
Leakage Rate Testing Program. The proposed amendment does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. The primary containment
function is to provide an essentially leak tight barrier against the
uncontrolled release of radioactivity to the environment for
postulated accidents. As such, the containment and the testing
requirements to periodically demonstrate the integrity of the
containment exist to ensure the plant's ability to mitigate the
consequences of an accident, and do not involve any accident
precursors or initiators.
Therefore, the probability of occurrence of an accident
previously evaluated is not significantly increased by the proposed
amendment.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for development of the NAPS performance-based
testing program. Implementation of these guidelines continues to
provide adequate assurance that during design basis accidents, the
primary containment and its components will limit leakage rates to
less than the values assumed in the plant safety analyses. The
potential consequences of extending the ILRT [integrated leak rate
test] interval to 15 years have been evaluated by analyzing the
resulting changes in risk. The increase in risk in terms of person-
rem per year within 50 miles resulting from design basis accidents
was estimated to be acceptably small and determined to be within the
guidelines published in RG 1.174 [``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific changes to the Licensing Basis'']. Additionally, the
proposed change maintains defense-in-depth by preserving a
reasonable balance among prevention of core damage, prevention of
containment failure, and consequence mitigation. NAPS has determined
that the increase in Conditional Containment Failure Probability due
to the proposed change is very small.
Therefore, it is concluded that the proposed amendment does not
significantly increase the consequences of an accident previously
evaluated.
Based on the above discussion, it is concluded that the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--Does the change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for the development of the NAPS performance-
based leakage testing program, and establishes a 15-year interval
for the performance of the containment ILRT. The containment and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident, do not involve any accident precursors
or initiators. The proposed change does not involve a physical
change to the plant (i.e., no new or different type of equipment
will be installed) or a change to the manner in which the plant is
operated or controlled.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
Criterion 3--Does this change involve a significant reduction in
a margin of safety?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for the development of the NAPS performance-
based leakage testing program, and establishes a 15-year interval
for the performance of the containment ILRT. This amendment does not
alter the manner in which safety limits, limiting safety system
setpoints, or limiting conditions for operation are determined. The
specific requirements and conditions of the Containment Leakage Rate
Testing Program, as defined in the TS, ensure that the degree of
primary containment structural integrity and leak-tightness that is
considered in the plant's safety analysis is maintained. The overall
containment leakage rate limit specified by the TS is maintained,
and the Type A, Type B, and Type C containment leakage tests will be
performed at the frequencies established in accordance with the NRC-
accepted guidelines of NEI 94-01, Revision 3-A.
Containment inspections performed in accordance with other plant
programs serve to provide a high degree of assurance that the
containment will not degrade in a manner that is not detectable by
an ILRT. A risk assessment using the current NAPS PRA [probabilistic
risk assessment] model concluded that extending the ILRT test
interval from 10 years to 15 years results in a small change to the
NAPS risk profile.
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 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: Robert Pascarelli.
III. 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.
[[Page 52071]]
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 can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
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 application for amendments: December 16, 2011, as
supplemented by letters dated January 20, March 1, March 16, April 18,
July 11, July 20, August 31, and November 2, 2012; April 5, June 28,
August 7, and December 18, 2013; and February 14, April 3, April 11,
and July 24, 2014.
Brief description of amendments: The amendments revised the
Technical Specifications and the Updated Final Safety Analysis Report
to add the new Protected Service Water (PSW) System to the plant's
licensing basis as an additional method of achieving and maintaining
safe shutdown of the reactors in the event of a high-energy line break
or a fire in the turbine building, which is shared by all three units.
Date of Issuance: August 13, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 386, 388, and 387. A publicly-available version is
in ADAMS under Accession No. ML14206A790. Documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the license and the TSs.
Date of initial notice in Federal Register: July 10, 2012, (77 FR
40652). The supplemental letters dated January 20, March 1, March 16,
April 18, July 11, July 20, August 31, and November 2, 2012; April 5,
June 28, August 7, and December 18, 2013; and February 14, April 3,
April 11, and July 24, 2014, provided additional information that
clarified the application, did not expand the scope of the application
as noticed, and did not change the staff's 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 August 13, 2014.
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: August 5, 2013, as supplemented by
letter dated January 28, 2014.
Brief description of amendment: The amendment revised the
structural design basis related to the leak-before-break analysis for
the reactor coolant system piping described in Section 4.3.6 of the
Updated Safety Analysis Report.
Date of issuance: August 7, 2014.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 276. A publicly-available version is in ADAMS under
Accession No. ML14209A027; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-40: The amendment
revised the design basis as described in the Updated Safety Analysis
Report.
Date of initial notice in Federal Register: April 8, 2014 (79 FR
19400).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated August 7, 2014.
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 application for amendment: August 20, 2012, as supplemented
by letters dated October 25, 2012, November 8, 2012, July 2, 2013, and
June 16, 2014.
Brief description of amendments: The amendments revise the
condensate storage tank level requirement specified in Technical
Specification surveillance requirement 3.7.6.1.
Date of issuance: August 15, 2014.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1--195, Unit 2--191. A publicly-available
version is in ADAMS under Accession No. ML14155A302; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-2 and NPF-8: The amendments
revised the Renewed Facility Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: January 15, 2013 (78 FR
3037). The supplemental letters dated October 25, 2012, November 8,
2012, July 2, 2013, and June 16, 2014, 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 August 15, 2014.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia and
Docket No. 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry
County, Virginia
Date of application for amendment: June 26, 2013, as supplemented
by letter dated January 23, 2014.
Brief description of amendment: The license amendments approve the
generic application of Appendix D, ``Qualification of the ABB-NV and
WLOP Critical Heat Flux (CHF) Correlations in the Dominion VIPRE-D
Computer Code,'' to Fleet Report DOM-NAF-2-A, ``Reactor Core Thermal-
Hydraulics Using the VIPRE-D Computer Code,'' the plant-specific
applications of Appendix D to Fleet Report DOM-NAF-2-A to North Anna
and Surry Power Stations, an added Surry reactor core safety limit, an
increase in the Surry Minimum Temperature for Criticality (MTC), and
modified references to MTC.
Date of issuance: August 12, 2014.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 271, 253, 283, and 283. A publicly-available
version is in ADAMS under Accession No. ML14169A359. Documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendment.
Renewed Facility Operating License Nos. NPF-4 and NPF-7, DPR-32 and
DPR-37: Amendments changed the licenses.
Date of initial notice in Federal Register: September 3, 2013 (78
FR 54292). The supplemental dated January 23, 2014, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed,
[[Page 52072]]
and did not change the staffs original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 12, 2014.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 23, 2013.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.6.5, ``CORE OPERATING LIMITS REPORT (COLR),'' to
replace the methodology of Westinghouse Electric Company LLC topical
report WCAP-11596-P-A, ``Qualification of the Phoenix-P/ANC Nuclear
Design System for Pressurized Water Reactor Cores,'' with WCAP-16045-P-
A, ``Qualification of the Two-Dimensional Transport Code PARAGON,'' and
WCAP-16045-P-A, Addendum 1-A, ``Qualification of the NEXUS Nuclear Data
Methodology,'' to determine core operating limits.
Date of issuance: August 7, 2014.
Effective date: As of its date of issuance and shall be implemented
prior to core reload during Refueling Outage 20, currently expected to
begin in January 2015.
Amendment No.: 209. A publicly-available version is in ADAMS under
Accession No. ML14156A246; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 10, 2013 (78
FR 74186).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 7, 2014.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: January 23, 2014. A redacted version was
provided by letter dated March 31, 2014.
Brief description of amendment: The amendment revised the Cyber
Security Plan Implementation Milestone No. 8 completion date and the
physical protection license condition.
Date of issuance: August 14, 2014.
Effective date: As of its date of issuance and shall be implemented
within 90 days.
Amendment No.: 210. A publicly-available version is in ADAMS under
Accession No. ML14209A023; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License.
Date of initial notice in Federal Register: June 6, 2014 (79 FR
32765).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 14, 2014.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 22nd day of August 2014.
For the Nuclear Regulatory Commission.
A. Louise Lund,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2014-20671 Filed 8-29-14; 8:45 am]
BILLING CODE 7590-01-P