[Federal Register Volume 81, Number 30 (Tuesday, February 16, 2016)]
[Notices]
[Pages 7835-7847]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-02916]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0026]
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 (AEA), 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 January 16, 2016, to February 1, 2016. The
last biweekly notice was published on February 2, 2016.
DATES: Comments must be filed by March 17, 2016. A request for a
hearing must be filed by April 18, 2016.
ADDRESSES: You may submit comments by any of the following methods
(unless
[[Page 7836]]
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-2016-0026. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, 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: Paula Blechman, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-2242, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0026 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-2016-0026.
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 of this document.
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-2016-0026, facility name, unit
number(s), application date, and subject in your comment submission.
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, (2) create the possibility of a new or different
kind of accident from any accident previously evaluated, or (3) involve
a significant reduction in a margin of safety. The basis for this
proposed determination for each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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 within 60 days, 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
[[Page 7837]]
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 set forth 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 with respect to resolution of that person's admitted
contentions, including the opportunity to present evidence and to
submit a cross-examination plan for cross-examination of witnesses,
consistent with NRC regulations, policies and procedures.
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).
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, 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 unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission by April
18, 2016. The petition must be filed in accordance with the filing
instructions in the ``Electronic Submissions (E-Filing)'' section of
this document, and should meet the requirements for petitions for leave
to intervene set forth in this section, except that under Sec.
2.309(h)(2) a State, local governmental body, or Federally-recognized
Indian Tribe, or agency thereof does not need to address the standing
requirements in 10 CFR 2.309(d) if the facility is located within its
boundaries. A State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may also have the opportunity to
participate under 10 CFR 2.315(c).
If a hearing is granted, any person who does not wish, or is not
qualified, to become a party to the proceeding may, in the discretion
of the presiding officer, be permitted to make a limited appearance
pursuant to the provisions of 10 CFR 2.315(a). A person making a
limited appearance may make an oral or written statement of position on
the issues, but may not otherwise participate in the proceeding. A
limited appearance may be made at any session of the hearing or at any
prehearing conference, subject to the limits and conditions as may be
imposed by the presiding officer. Persons desiring to make a limited
appearance are requested to inform the Secretary of the Commission by
April 18, 2016.
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 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/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
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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, in some instances, 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 these license amendment
applications, 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.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2 (ANO-2), Pope County, Arkansas
Date of amendment request: December 22, 2015. A publicly-available
version is in ADAMS under Accession No. ML15356A657.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) to add a short Allowed Outage Time (AOT)
to restore an inoperable system for conditions under which the existing
specifications require a plant shutdown. The proposed amendment is
consistent with the NRC-approved Technical Specifications Task Force
(TSTF) change traveler 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.'' The availability of
TSTF-426, Revision 5, was published in the Federal Register on May 30,
2013 (78 FR 32476). The AOT would be added to specifications governing
the pressurizer heaters, containment spray trains, and control room
emergency air conditioning and ventilation systems. In addition to the
scope of the TSTF-426 TSs revisions, the amendment would add a TS
Action to address a single pressurizer proportional heater group having
a capacity of less than 150 kilowatts.
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 provides a short AOT to restore an
inoperable system for conditions under which the existing TSs
require a plant shutdown to begin within one hour in accordance with
LCO 3.0.3. In addition, a new TS Action associated with Pressurizer
proportional heater capacity for a single proportional heater group
is proposed. Entering into TS 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 AOTs are no different from the
[[Page 7839]]
consequences of the same accident during the existing one-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 kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. 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. The proposed change does 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 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
analyses 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 this 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 one hour is not
significant.
The new Pressurizer proportional heater capacity Action permits
72 hours to restore the affect heater group to an operable status,
consistent with the STS [Standard TSs] and consistent with TS
requirements associated with single train inoperabilities. The
proportional heaters are not credited in the ANO-2 accident
analyses, but aid in Pressurizer pressure control during a loss of
offsite power event that results in the need to perform a natural
circulation cool down of the plant. The associated STS bases for the
standard 72-hour AOT assumes [that] the likelihood of a loss of
offsite power event during this time period that would require a
demand on the proportional heaters is minimal and acknowledges the
use of non-vital powered backup heater groups absent a loss of
offsite power event. Note also that under emergency conditions, an
Emergency Diesel Generator or the Alternate AC [alternating current]
Diesel Generator (i.e., Station Blackout diesel) can be aligned to
power any of the non-vital Pressurizer backup heater groups. As a
result, the change to the margin of safety provided by the new 72-
hour AOT for a single proportional heater train 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: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Meena K. Khanna.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: October 15, 2015. A publicly-available
version is in ADAMS under Accession No. ML15301A765.
Description of amendment request: The amendments would revise the
St. Lucie Plant, Unit Nos. 1 and 2, Renewed Facility Operating
Licenses' licensing bases to allow the use of the commercially
available code ``Generation of Thermal-Hydraulic Information for
Containments (GOTHIC Version 7.2b(QA)),'' to model the containment
response following the inadvertent actuation of the containment spray
system during normal plant operation (referred to as the vacuum
analysis). The amendments would also update the licensing bases to
credit the design-basis ability of the containment vessel to withstand
a higher external pressure differential of 1.04 pounds per square inch
(psi) (1.05 psi for Unit No. 2), and will update Technical
Specification 3.6.1.4 for both units to revise the allowable
containment operating pressure range.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This proposed amendment is related to the analysis of the
maximum external pressure that the reactor containment building will
experience. A proposed change to the Technical Specifications will
limit the allowable external pressure during operation to a value
consistent with that considered in the analysis. The analysis is
being revised to consider containment spray pump flow higher than
previously considered. Containment spray pumps cool and depressurize
the containment building; therefore, higher flow impacts the
analysis of external pressure on the containment building. The
proposed amendment is for the use of a different analysis
methodology using the GOTHIC computer code instead of the A-TEMPT
and WATEMPT codes that were originally used for the Unit 1 and Unit
2 analyses respectively. The original codes are not currently
available. The GOTHIC code is an accepted code for similar analysis.
The analysis performed demonstrates that in the postulated event of
an inadvertent start of two containment spray pumps, the loading the
reactor containment building will experience is within the design of
the structure. With this load, the stresses experienced by the
reactor containment building remain below the code allowable
stresses.
The probability of occurrence of an event that would expose the
containment building to external pressure is not increased by the
change in the analysis methodology used. The probability of the
initiating event, inadvertent start of both containment spray pumps,
is unchanged.
The consequences of an event where the containment building is
exposed to external pressure will not be increased as the resulting
external pressure on the containment vessel remains within the
design, which provides a large margin to the buckling pressure.
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.
This proposed amendment changes the methodology for analyzing an
event that results in exposing the reactor containment vessel to
external pressure. A proposed change to the Technical Specifications
will limit the external pressure during operation to a value
consistent with the initial condition considered in the analysis.
The potential for a new or different kind of accident is not created
by the use of a different analysis methodology for a previously
defined event.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This proposed amendment changes the methodology for analyzing an
event that results in exposing the reactor containment building to
external pressure. A proposed change to the Technical Specifications
will limit the allowable external pressure during operation to a
value consistent with the starting point considered in the analysis.
The technical evaluation demonstrates that the use of the GOTHIC
computer code to determine maximum containment external pressure
will result in realistic results similar
[[Page 7840]]
to the original analysis with the A-TEMPT and WATEMPT codes. The
margin of safety in this analysis is maintained by assuring the
resulting external pressure acting on the reactor containment vessel
maintains significant margin to the buckling pressure in accordance
with Section III of the ASME [American Society of Mechanical
Engineers] code. For Unit 2, the original code of record limited the
maximum external pressure to \1/3\ of the expected buckling
pressure. The analysis of the increased external pressure for Unit 2
has been performed in accordance with the original code of record.
The original code of record for Unit 1 was under development at the
time and made reference to ASME Section VIII for the analysis of
external pressure. The rules of ASME Section VIII at that time
limited the maximum external pressure to \1/4\ of the expected
buckling pressure. In order to increase the allowable external
pressure, the analysis of external pressure was performed using a
later version of the ASME code which allows a maximum external
pressure of \1/3\ of the buckling pressure. The later version of the
code used for Unit 1 uses a methodology for determining the maximum
external pressure consistent with the code used for Unit 2.
Although the margin between the allowable external pressure and
the expected buckling pressure for Unit 1 will be changed from a
factor of 4 to a factor of 3, substantial margin is maintained in
accordance with more current versions of ASME III.
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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
NRC Branch Chief: Benjamin G. Beasley.
South Carolina Electric and Gas Company 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: December 17, 2015. A publicly-available
version is in ADAMS under Accession No. ML15351A165.
Description of amendment request: The proposed change, if approved,
would amend Combined License (COL) Nos. NPF-93 and NPF-94 for VCSNS.
The requested amendment proposes to rename, relocate, and add radiation
detectors to provide monitoring of the radiologically controlled area
ventilation system (VAS) exhaust from the radiologically controlled
areas of the auxiliary building and annex building. The changes in the
proposed amendment are located primarily in the VCSNS Updated Final
Safety Analysis Report (UFSAR) Tier 2 information, and involve require
conforming changes to COL Appendix C, ``Inspections, Tests, Analyses,
and Acceptance Criteria,'' and departing from certified AP1000 Design
Control Document (DCD) Tier 1 information. 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
52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design functions of the VAS include prevention of the
unmonitored release of airborne radioactivity to the atmosphere or
adjacent plant areas by providing monitoring of the VAS exhaust from
radiologically controlled areas of the auxiliary building and annex
building, and to automatically isolate the selected building areas
and start the containment air filtration system (VFS) upon detection
of high radioactivity. The proposed changes to the VAS to relocate
and add radiation detectors are acceptable as they maintain these
design functions. These proposed changes to the VAS design as
described in the current licensing basis do not have an adverse
effect on any of the design functions of the systems. The proposed
changes do not affect the support, design, or operation of
mechanical and fluid systems required to mitigate the consequences
of an accident. There is no change to plant systems or the response
of systems to postulated accident conditions. There is no change to
the predicted radioactive releases due to postulated accident
conditions. The plant response to previously evaluated accidents or
external events is not adversely affected, nor do the proposed
changes described create any new accident precursors.
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 changes revise the VAS design as described in the
current licensing basis to enable the system to perform required
design functions, and are consistent with other UFSAR information.
The proposed changes do not change the design requirements for the
system. The relocated and new VAS radiation detectors are designed
to the same equipment specifications, including required sensitivity
and range, as the existing radiation detectors. The relocated and
new VAS radiation detectors monitor the same parameters, as well as
perform the same design functions, as the existing radiation
detectors. The proposed changes to the system do not result in a new
failure mechanism or introduce any new accident precursors. No
design function described in the UFSAR is adversely affected by the
proposed changes. The proposed changes do not result in a new
failure mode, malfunction or sequence of events that could affect
safety or safety-related equipment. The proposed changes do not
allow for a new fission product release path, result in a new
fission product barrier failure mode, or create a new sequence of
events that would result in significant fuel cladding failures.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not change the codes or standards for
the radiation detectors, or functionality of the ductwork in the
auxiliary building and annex building. The proposed changes have no
adverse effect on the nonsafety-related system design functions of
the VAS for the prevention of the unmonitored release of airborne
radioactivity to the atmosphere or adjacent plant areas by providing
monitoring of the VAS exhaust from radiologically controlled areas
of the auxiliary building and annex building, and to automatically
isolate the selected building areas and start the VFS upon detection
of high radioactivity. The proposed changes do not affect safety-
related equipment or equipment whose failure could initiate an
accident. The proposed changes to relocate and add radiation
detectors do not adversely interface with safety-related equipment
or fission product barriers. Therefore, the proposed changes do not
affect any safety-related equipment, design code, function, design
analysis, safety analysis input or result, or design/safety margin.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the requested changes, thus, no margin of
safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
[[Page 7841]]
Acting NRC Branch Chief: John McKirgan.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: December 22, 2015. A publicly-available
version is in ADAMS. under Accession No.ML15356A656.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91 and NPF-92 for VEGP, Units 3 and 4,
respectively. The requested amendment proposes to depart from approved
AP1000 Design Control Documents (DCD) Tier 2 information (text, tables,
and figures) and involved Tier 2* information (as incorporated into the
Updated Final Safety Analysis Report (UFSAR) as plant specific DCD
information), and also involves a change to a license condition.
Specifically, the requested amendment proposes changes to the design of
auxiliary building Wall 11 and proposes other changes to the licensing
basis for use of Seismic Category II structures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect the operation of
any systems or equipment inside or outside the auxiliary building
that could initiate or mitigate abnormal events, e.g., accidents,
anticipated operational occurrences, earthquakes, floods, tornado
missiles, and turbine missiles, or their safety or design analyses,
evaluated in the UFSAR. The changes do not adversely affect any
design function of the auxiliary building or the systems and
equipment contained therein. The ability of the affected auxiliary
building [Main Steam Isolation Valve] MSIV compartments to withstand
the pressurization effects from the design basis pipe rupture is not
adversely affected by the removal of the Wall 11 upper vent
openings, because vents at these locations are not credited in the
subcompartment pressurization analysis. MSIV compartment
temperatures following the limiting one square foot pipe rupture
with the vent openings removed remain acceptably within the envelope
for environmental qualification of equipment in the compartments.
The credit of seismic Category II Wall 11.2 as a [high energy line
break] HELB barrier and the seismic Category II turbine building
first bay and associated missile barriers to protect Wall 11
openings from tornado missiles continues to provide adequate
protection of structures, systems, and components (SSCs) required to
safely shut down the plant, as these structures are designed to the
same requirements as seismic Category I structures, and with the
additional HELB loadings assumed, remain well within the applicable
acceptance criteria.
Therefore, the proposed amendment does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not change the design function of the
auxiliary building or of any of the systems or equipment in the
auxiliary building or elsewhere within the Nuclear Island structure.
These proposed changes do not introduce any new equipment or
components that would result in a new failure mode, malfunction or
sequence of events that could affect safety-related or nonsafety-
related equipment. This activity will not allow for a new fission
product release path, result in a new fission product barrier
failure mode, or create a new sequence of events that would result
in significant fuel cladding failures.
Therefore, this activity does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety for the design of the auxiliary building is
maintained through continued use of the current codes and standards
as stated in the UFSAR and adherence to the assumptions used in the
analyses of this structure and the events associated with this
structure. The auxiliary building will continue to maintain a
seismic Category I rating which preserves the current structural
safety margins. The 3-hour fire rating requirements for the impacted
auxiliary building walls are maintained. The Wall 11 upper vents are
not credited in the subcompartment pressurization analysis and the
remaining vents and pressure relief devices provide sufficient
venting to maintain the MSIV compartment pressures below the design
limit and design basis. The credit of turbine building Wall 11.2 as
a HELB barrier provides protection of Wall 11 from selected dynamic
effects, which in turn provides that essential SSCs remain protected
from the effects of postulated HELB events. The credit of the
seismic Category II turbine building first bay and associated
missile barriers to provide protection of Wall 11 openings from
tornado missiles provides sufficient protection for the essential
SSCs located in the auxiliary building in the vicinity of Wall 11
from the effects of external missiles. Thus, the requested changes
will not adversely affect any safety-related equipment, design code,
function, design analysis, safety analysis input or result, or
design/safety margin. No safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by the requested change,
thus, no margin of safety is reduced.
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. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
Acting NRC Branch Chief: John McKirgan.
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, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: December 15, 2015. A publicly-available
version is in ADAMS under Accession No. ML15351A023.
Description of amendment request: The amendments would modify the
Technical Specifications (TSs) to risk-inform the requirements
regarding selected Required Action end states by incorporating TS Task
Force (TSTF) traveler TSTF-423, Revision 1, ``Technical Specification
End States, NEDC-32988-A.'' Additionally, it would modify the TS
Required Actions with a Note prohibiting the use of limiting condition
for operation 3.0.4.a when entering the preferred end state (Mode 3) on
startup. The Notice of Availability for TSTF-423, Revision 1, was
published in the Federal Register on February 18, 2011 (76 FR 9614).
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 allows a change to certain required end
states when the TS Completion Times for remaining in power operation
will be exceeded. Most of the requested technical specification (TS)
changes are to permit an end state of hot shutdown (Mode 3) rather
than an end state of cold shutdown (Mode 4) contained in the current
TS. The request was limited to: (1)
[[Page 7842]]
Those end states where entry into the shutdown mode is for a short
interval, (2) entry is initiated by inoperability of a single train
of equipment or a restriction on a plant operational parameter,
unless otherwise stated in the applicable TS, and (3) the primary
purpose is to correct the initiating condition and return to power
operation as soon as is practical. Risk insights from both the
qualitative and quantitative risk assessments were used in specific
TS assessments.
Such assessments are documented in Section 6 of topical report
NEDC-32988-A, Revision 2, ``Technical Justification to Support Risk
Informed Modification to Selected Required Action End States for BWR
Plants.'' They provide an integrated discussion of deterministic and
probabilistic issues, focusing on specific TSs, which are used to
support the proposed TS end state and associated restrictions. The
NRC staff finds that the risk insights support the conclusions of
the specific TS assessments. Therefore, the probability of an
accident previously evaluated is not significantly increased, if at
all. The consequences of an accident after adopting TSTF-423 are no
different than the consequences of an accident prior to adopting
TSTF-423. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns.
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 does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded (i.e., entry into hot shutdown rather
than cold shutdown to repair equipment) will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change and the commitment by the licensee to adhere to the guidance
in TSTF-IG-05-02, ``Implementation Guidance for TSTF-423, Revision
1, `Technical Specifications End States, NEDC-32988-A,' '' will
further minimize possible concerns.
Thus, based on the above, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The Boiling Water Reactor Owners' Group's risk
assessment approach is comprehensive and follows NRC staff guidance
as documented in Regulatory Guides (RG) 1.174 and 1.177. In
addition, the analyses show that the criteria of the three-tiered
approach for allowing TS changes are met. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A risk assessment was performed to justify
the proposed TS changes. The net change to the margin of safety is
insignificant.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based upon the reasoning presented above, SNC concludes that the
requested change involves no significant hazards consideration, as
set forth in 10 CFR 50.92(c), ``Issuance of Amendment.''
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: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Iverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: November 20, 2015, as supplemented by
letter dated January 12, 2016. Publicly-available versions are in ADAMS
under Accession Nos. ML15324A297 and ML16012A457, respectively.
Description of amendment request: The proposed change would revise
the setpoint requirements in Technical Specification (TS) 3.3.5, ``Loss
of Power Diesel Generator Start Instrumentation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed license amendment request changes the TS 3.3.5
requirements for loss of power diesel generator start
instrumentation to enable elimination of manual actions for
protection of safety-related equipment from degraded voltage
conditions during design basis events. Elimination of these manual
actions is required to fulfill an existing License Condition on each
unit.
The proposed change increases the Allowable Value (AV) for the
4.16 kV Emergency Bus Degraded Grid Voltage Actuation function.
Installation of new, higher precision Degraded Voltage Relays (DVRs)
makes possible an increase in the DVR actuation setpoint
(encompassed by the AV) to a level which provides fully automatic
protection of safety-related equipment while minimizing the chance
of unwanted disconnection from the preferred offsite power source,
which is itself an analyzed condition.
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 license change request changes the TS 3.3.5
requirements for loss of power diesel generator start
instrumentation to enable elimination of manual actions for
protection of safety-related equipment from degraded voltage
conditions during design basis events. Elimination of these manual
actions is required to fulfill an existing License Condition on each
unit.
The proposed changes to TS 3.3.5 do not change the methods of
normal plant operation nor the methods of response to transient
conditions, save that the range of automatic action provided by the
DVRs is expanded. This change will eliminate the need for manual
action from the degraded voltage protection scheme, as required by a
License Condition for each unit, to achieve compliance with 10 CFR
50.55a(h)(2) and 10 CFR part 50, Appendix A, General Design
Criterion 17--Electric Power Systems.
Accordingly, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is provided by the performance capability of
plant equipment in preventing or mitigating challenges to fission
product barriers under postulated operational transient and accident
conditions. Since the proposed license amendment request changes the
TS 3.3.5 requirements for loss of power diesel generator start
instrumentation to enable elimination of manual actions for
protection of safety-related equipment from degraded voltage
conditions during design basis events, it will tend to increase the
margin of safety by better protecting the safety-related plant
equipment.
Based on the above, 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.
[[Page 7843]]
Attorney for licensee: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Iverness Center
Parkway, Birmingham, AL 35201.
NRC Branch Chief: Michael T. Markley.
STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499,
South Texas Project (STP), Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 19, 2013, as supplemented by
letters dated October 3, October 31, November 13, November 21, and
December 23, 2013 (two letters); January 9, February 13, February 27,
March 17, March 18, May 15, May 22, June 25, and July 15, 2014; and
March 10, March 25, and August 20, 2015. For the convenience of the
reader, the ADAMS accession numbers of the amendment request,
supplements, and additional documents (if publicly available) are
provided below in a table in the ``Availability of Documents'' section.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) and licensing basis for Facility
Operating License Nos. NPF-76 and NPF-80, for STP, Units 1 and 2, as
documented in the Updated Final Safety Analysis Report (UFSAR). The
changes incorporate use of both a deterministic and a risk-informed
approach to address safety issues discussed in Generic Safety Issue
(GSI)-191, ``Assessment of Debris Accumulation on PWR [Pressurized-
Water Reactor] Sump Performance,'' and to close Generic Letter (GL)
2004-02, ``Potential Impact of Debris Blockage on Emergency
Recirculation during Design Basis Accidents at Pressurized-Water
Reactors,'' dated September 13, 2004 (ADAMS Accession No. ML042360586),
for STP.
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 changes are a methodology change for assessment of
debris effects that adds the results of a risk-informed evaluation
to the STP licensing basis, changes to the [emergency core cooling
system (ECCS)] and [containment spray system (CSS)] TS to extend the
required completion time for potential [loss-of-coolant accident
(LOCA)] debris related effects and associated administrative TS
changes. The methodology change concludes that the ECCS and CSS will
have sufficient defense-in-depth and safety margin and will operate
with high probability following a LOCA when considering the impacts
and effects of debris accumulation on containment emergency sump
strainers in recirculation mode, as well as core flow blockage due
to in-vessel effects, following loss of coolant accidents. The
methodology change also supports the changes to the TS.
There is no significant increase in the probability of an
accident previously evaluated. The proposed changes address
mitigation of loss of coolant accidents and have no effect on the
probability of the occurrence of a loss of coolant accident. The
proposed methodology and TS changes do not implement any physical
changes to the facility or any [structures, systems, and components
(SSCs)], and do not implement any changes in plant operation that
could lead to a different kind of accident.
The proposed changes do not involve a significant increase in
the consequences of an accident previously evaluated. The
methodology change confirms that required SSCs supported by the
containment sumps will perform their safety functions with a high
probability, as required, and does not alter or prevent the ability
of SSCs to perform their intended function to mitigate the
consequences of an accident previously evaluated within the
acceptance limits. The safety analysis acceptance criteria in the
UFSAR continue to be met for the proposed methodology change. The
evaluation of the changes determined that containment integrity will
be maintained. The dose consequences were considered in the
assessment and quantitative evaluation of the effects on dose using
input from the risk-informed approach shows the increase in dose
consequences is small.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any the accident
previously evaluated in the UFSAR.
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 changes are a methodology change for assessment of
debris effects from LOCAs that are already evaluated in the STP
UFSAR, an extension of TS required completion time for potential
LOCA debris related effects on ECCS and CSS, and associated
administrative changes to the TS. No new or different kind accident
is being evaluated. None of the changes install or remove any plant
equipment, or alter the design, physical configuration, or mode of
operation of any plant structure, system or component. The proposed
changes do not introduce any new failure mechanisms or malfunctions
that can initiate an accident. The proposed changes do not introduce
failure modes, accident initiators, or equipment malfunctions that
would cause a new or different kind of accident.
Therefore, the proposed changes do not create the possibility
for 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 proposed changes are a methodology change for assessment of
debris effects from LOCAs that are already evaluated in the STP
UFSAR, an extension of TS required completion time for potential
LOCA debris related effects on ECCS and CSS, and associated
administrative changes to the TS. The effects from a full spectrum
of LOCAs, including double-ended guillotine breaks for all piping
sizes up to and including the largest pipe in the reactor coolant
system, are analyzed. Appropriate redundancy and consideration of
loss of offsite power and worst case single failure are retained,
such that defense-in-depth is maintained.
Application of the risk-informed methodology showed that the
increase in risk from the contribution of debris effects is very
small as defined by [NRC Regulatory Guide (RG) 1.174, ``An Approach
for Using Probabilistic Risk Assessment in Risk-Informed Decisions
on Plant-Specific Changes to the Licensing Basis''] and that there
is adequate defense in depth and safety margin. Consequently, STP
determined that the risk-informed method demonstrates the
containment sumps will continue to support the ability of safety
related components to perform their design functions when the
effects of debris are considered. The proposed change does not alter
the manner in which safety limits are determined or acceptance
criteria associated with a safety limit. The proposed change does
not implement any changes to plant operation, and does not
significantly affect SSCs that respond to safely shutdown the plant
and to maintain the plant in a safe shutdown condition. The proposed
change does not significantly affect the existing safety margins in
the barriers for the release of radioactivity. There are no changes
to any of the safety analyses in the UFSAR.
Defense in depth and safety margin was extensively evaluated for
the methodology change and the associated TS changes. The evaluation
determined that there is substantial defense in depth and safety
margin that provide a high level of confidence that the calculated
risk for the methodology and TS changes is conservative and that the
actual risk is likely much lower.
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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Availability of Documents
For further details with respect to this action, see the
application for license amendment dated June 19, 2013, listed below in
the table, in addition to supplements, requests for additional
information responses, and other relevant documents.
[[Page 7844]]
------------------------------------------------------------------------
Title Date ADAMS Accession No.
------------------------------------------------------------------------
SECY-12-0093, ``Closure Options for 07/09/2012 ML121320270
Generic Safety Issue-191,
Assessment of Debris Accumulation
on Pressurized-Water Reactor Sump
Performance.''
STP Pilot Submittal and Request for 01/31/2013 ML13043A013
Exemption for a Risk-Informed
Approach to Resolve Generic Safety
Issue (GSI)-191.
NRC Letter to STPNOC, ``South Texas 04/01/2013 ML13066A519
Project, Units 1 and 2--
Supplemental Information Needed
for Acceptance of Requested
Licensing Action Re: Request for
Exemption for a Risk-Informed
Approach to Resolve Generic Safety
Issue 191''.
Revised STP Pilot Submittal and 06/19/2013 ML131750250
Requests for Exemptions and (package)
License Amendment for a Risk-
Informed approach to Resolving
Generic Safety Issue (GSI)-191.
NRC Letter to STPNOC, ``South Texas 08/13/2013 ML13214A031
Project, Units 1 and 2--Acceptance
of Requests for Exemptions and
License Amendment Request for
Approval of a Risk-Informed
Approach to Resolve Generic Safety
Issue GSI-191''.
Corrections to Information Provided 10/03/2013 ML13295A222
in Revised STP Pilot Submittal and
Requests for Exemptions and
License Amendment for a Risk-
Informed Approach to Resolving
Generic Safety Issue (GSI)-191.
Submittal of GSI-191 Chemical 10/31/2013 ML13323A673
Effects Test Reports. (package)
Supplement 1 to Revised STP Pilot 11/13/2013 ML13323A128
Submittal and Requests for (package)
Exemptions and License Amendment
for a Risk-Informed Approach to
Resolving Generic Safety Issue
(GSI)-191.
Supplement 1 to Revised STP Pilot 11/21/2013 ML13338A165
Submittal for a Risk-Informed
Approach to Resolving Generic
Safety Issue (GSI)-191 to
Supersede and Replace the Revised
Pilot Submittal.
Response to STP-GSI-191-EMCB-RAI-1. 12/23/2013 ML14015A312
Response to NRC Request for 12/23/2013 ML14015A311
Reference Document for STP Risk-
Informed GSI-191 Application.
Response to Request for Additional 01/09/2014 ML14029A533
Information re Use of RELAP5 in
Analyses for Risk-Informed GSI-191
Licensing Application.
Submittal of CASA Grande Code and 02/13/2014 ML14052A110
Analyses for STP's Risk-Informed (package, portions
GSI-191 Licensing Application. redacted)
Submittal of GSI-191 Chemical 02/27/2014 ML14072A075
Effects Test Reports. (package)
Response to NRC Accident Dose 03/17/2014 ML14086A383
Branch Request for Additional (package)
Information Regarding STP Risk-
Informed GSI-191 Application.
Submittal of CASA Grande Source 03/18/2014 (proprietary,
Code for STP's Risk-Informed GSI- non-public)
191 Licensing Application.
Second Submittal of CASA Grande 05/15/2014 ML14149A354
Source Code for STP's Risk-
Informed GSI-191 Licensing
Application.
First Set of Responses to April, 05/22/2014 ML14149A439
2014, Requests for Additional (package)
Information Regarding STP Risk-
Informed GSl-191 Licensing
Application--Revised.
Second Set of Responses to April, 06/25/2014 ML14178A467
2014, Requests for Additional (package)
Information Regarding STP Risk-
Informed GSI-191 Licensing
Application.
Third Set of Responses to April, 07/15/2014 ML14202A045
2014, Requests for Additional
Information Regarding STP Risk-
Informed GSI-191 Licensing
Application.
Submittal of Updated CASA Grande 03/10/2015 ML15072A092
Input for STP's Risk-Informed GSI-
191 Licensing Application.
Description of Revised Risk- 03/25/2015 ML15091A440
Informed Methodology and Responses
to Round 2 Requests for Additional
Information Regarding STP Risk-
Informed GSI-191 Licensing
Application.
Supplement 2 to STP Pilot Submittal 08/20/2015 ML15246A125
and Requests for Exemptions and
License Amendment for a Risk-
Informed Approach to Address
Generic Safety Issue (GSI)-191 and
Respond to Generic Letter (GL)
2004[dash]02.
------------------------------------------------------------------------
Attorney for licensee: Steve Frantz, Esq., Morgan, Lewis & Bockius,
1111 Pennsylvania Avenue NW., Washington, DC 20004.
NRC Branch Chief: Robert J. Pascarelli.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant
(WBN), Unit 2, Rhea County, Tennessee
Date of amendment request: December 15, 2015. A publicly-available
version is in ADAMS under Accession No. ML15362A023.
Description of amendment request: The amendment would revise
Technical Specifications (TSs) 3.4.17, ``Steam Generator (SG) Tube
Integrity''; 5.7.2.12, ``Steam Generator (SG) Program''; and 5.9.9,
``Steam Generator Tube Inspection Report,'' to exclude portions of the
SG tubes below the top of the tube sheet from needing to be plugged.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
Allowing the use of an alternate repair criteria as proposed in
this amendment request does not involve a significant increase in
the probability or consequence of an accident previously evaluated.
The presence of the tubesheet enhances the tube integrity in the
region of the hardroll by precluding tube deformation beyond its
initial expanded outside diameter. The resistance to both tube
rupture and tube collapse is strengthened by the presence of the
tubesheet in that region. Hardrolling of the tube into the tubesheet
results in an interference fit between the tube and the tubesheet.
Tube rupture cannot occur because the contact between the tube and
tubesheet does not permit sufficient movement of tube material. In a
similar manner, the tubesheet does not permit sufficient movement of
tube material to permit buckling collapse of the tube during
postulated loss-of-coolant-accident (LOCA) loadings.
The type of degradation for which the F* [the length of
mechanical expansion required to prevent pullout for all normal
operating and postulated accident conditions] has been developed
(cracking with a circumferential orientation) can theoretically lead
to a postulated tube rupture event, provided that the postulated
through-wall circumferential crack exists near the top of the
tubesheet. An evaluation including analysis and testing has
[[Page 7845]]
been performed to determine the resistive strength of roll expanded
tubes within the tubesheet. That evaluation provides the basis for
the acceptance criteria for tube degradation subject to the F*
criterion.
The F* length of roll expansion is sufficient to preclude tube
pullout from tube degradation located below the F* distance,
regardless of the extent of the tube degradation. The existing
technical specification leakage rate requirements and accident
analysis assumptions remain unchanged in the unlikely event that
significant leakage from this region does occur. As noted above,
tube rupture and pullout are not expected for tubes using the ARC
[alternative repair criterion]. Any leakage out of the tube from
within the tubesheet at any elevation in the tubesheet is fully
bounded by the existing Main Steam Line Break (MSLB) analysis
included in the WBN Unit 2 Final Safety Analysis Report (FSAR).
Therefore, the proposed ARC does not adversely impact any other
previously evaluated design basis accident.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Implementation of the proposed ARC does not introduce any
significant changes to the plant design basis. Use of the criterion
does not provide a mechanism to result in an accident initiated
outside of the region of the tubesheet expansion. A hypothetical
accident as a result of any tube degradation in the expanded portion
of the tube would be bounded by the existing tube rupture accident
analysis. Tube bundle structural integrity and leak tightness are
expected to be maintained.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The use of the ARC has been demonstrated to maintain the
integrity of the tube bundle commensurate with the requirements of
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR
[Pressurized-Water Reactor] Steam Generator Tubes,'' for indications
in the free span of tubes and the primary to secondary pressure
boundary under normal and postulated accident conditions. Acceptable
tube degradation for the F* criterion is any degradation indication
in the tubesheet region, more than the F* distance below either the
bottom of the transition between the roll expansion and the
unexpanded tube, or the top of the tubesheet, whichever is lower.
The safety factors used in the verification of the strength of the
degraded tube are consistent with the safety factors in the American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel
Code used in SG design. The F* distance has been verified by testing
to be greater than the length of roll expansion required to preclude
both tube pullout and significant leakage during normal and
postulated accident conditions. Resistance to tube pullout is based
upon the primary to secondary pressure differential as it acts on
the surface area of the tube, which includes the tube wall cross-
section, in addition to the inside diameter-based area of the tube.
The leak testing acceptance criteria are based on the primary to
secondary leakage limit in the technical specifications and the
leakage assumptions used in the UFSAR [Updated FSAR] accident
analyses. Implementation of the ARC will decrease the number of
tubes which must be taken out of service with tube plugs. Plugs
reduce the RCS flow margin; thus, implementation of the ARC will
maintain the margin of flow that would otherwise be reduced in the
event of increased plugging.
Based on the above, it is concluded that the proposed change
does not result in a significant reduction in or a loss of margin
with respect to plant safety as defined in the FSAR or the bases of
the WBN Unit 2 technical specifications.
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: Ralph E. Rodgers, General Counsel, Tennessee
Valley Authority, 400 West Summit Hill Dr., 6A West Tower, Knoxville,
TN 37902.
NRC Branch Chief: Benjamin G. Beasley.
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 (AEA), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit No. 2 (MPS2) and Unit No. 3 (MPS3), New
London County, Connecticut
Date of amendment request: January 15, 2015, as supplemented by
letters dated April 15, July 16, July 30, November 2, and December 1,
2015.
Brief description of amendment: The amendments revised the MPS2 and
MPS3 Technical Specifications (TSs) to adopt NRC-approved Technical
Specifications Task Force (TSTF) Standard Technical Specifications
(STS) Change Traveler TSTF-523, Revision 2, ``Generic Letter 2008-01,
Managing Gas Accumulation.''
Date of issuance: January 29, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 325 and 267. A publicly-available version is in
ADAMS under Accession No. ML16011A400; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-65 and NPF-49:
Amendments revised the Renewed Operating License and TSs.
Date of initial notice in Federal Register: July 21, 2015 (80 FR
43126). The supplemental letter dated April 15, 2015, was published
with the January 15, 2015, application, in the initial FR notice. The
supplemental letters dated July 16, July 30, November 2, and December
1, 2015, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a
[[Page 7846]]
Safety Evaluation dated January 29, 2016.
No significant hazards consideration comments received: No.
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: February 2, 2015, as supplemented by
letters dated August 11, 2015, and October 20, 2015.
Brief description of amendments: The amendments modified the
technical specifications (TSs) to allow for brief, inadvertent,
simultaneous opening of redundant secondary containment personnel
access doors during normal entry and exit conditions.
Date of issuance: January 28, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 220 and 182. A publicly-available version is in
ADAMS under Accession No. ML15356A140; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-39 and NPF-85:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: April 14, 2015 (80 FR
20022). The supplemental letters dated August 11, 2015, and October 20,
2015, 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 January 28, 2016.
No significant hazards consideration comments received: Yes.
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: February 23, 2015, as supplemented by
letters dated August 12, 2015, and October 20, 2015.
Brief description of amendments: The amendments modified the
technical specifications (TSs) to allow for brief, inadvertent,
simultaneous opening of redundant secondary containment personnel
access doors during normal entry and exit conditions.
Date of issuance: February 1, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendments Nos.: 303 and 307. A publicly-available version is in
ADAMS under Accession No. ML15350A179; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Renewed Facility Operating Licenses and the TSs.
Date of initial notice in Federal Register: April 14, 2015 (80 FR
20023). The supplemental letters dated August 12, 2015, and October 20,
2015, 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 February 1, 2016.
No significant hazards consideration comments received: Yes.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 15, 2015, as supplemented by
letters dated May 4, 2015, June 9, 2015, and January 12, 2016.
Brief description of amendment: The amendment revised the technical
specifications (TSs) to add a limiting condition for operation,
applicability, required actions, completion times, and surveillance
requirements for the residual heat removal containment spray and
associated interlock permissive instrumentation. A new TS Section
3.6.1.9, ``Residual Heat Removal (RHR) Containment Spray,'' has been
added to reflect the reliance on containment spray to maintain the
drywell within design temperature limits during a small steam line
break. In addition, the ``Drywell Pressure--High'' function that serves
as an interlock permissive to allow RHR containment spray mode
alignment has been relocated from the Technical Requirements Manual to
TS 3.3.5.1, ``Emergency Core Cooling System (ECCS) Instrumentation.''
Date of issuance: January 22, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 253. A publicly-available version is in ADAMS under
Accession No. ML15343A301; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-46: The amendment
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 17, 2015 (80 FR
13910). The supplemental letters dated May 4, 2015, June 9, 2015, and
January 12, 2016, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 22, 2016.
No significant hazards consideration comments received: No.
South Carolina Electric and Gas Company, 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: November 26, 2013, as supplemented by
letter dated June 3, 2015.
Brief description of amendment: The amendments are to Combined
License Nos. NPF-93 and NPF-94 for VCSNS, Units 2 and 3. The amendments
authorized changes to the VCSNS, Units 2 and 3, Updated Final Safety
Analysis Report to revise the details of the effective thermal
conductivity resulting from the oxidation of the inorganic zinc
component of the containment vessel coating system.
Date of issuance: October 9, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 34. A publicly-available version is in ADAMS under
Accession No.
ML15272A417; documents related to these amendments are listed in
the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses No. NPF-93 and NPF-94: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: February 19, 2014 (79
FR 9490). The supplemental letter dated June 3, 2015, 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.
[[Page 7847]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 9, 2015.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear
Plant (HNP), Unit No. 1, Appling County, Georgia
Date of application for amendment: September 1, 2015.
Brief description of amendments: The amendment revised the
Technical Specification value of the Safety Limit Minimum Critical
Power Ratio to support operation in the next fuel cycle.
Date of issuance: January 29, 2016.
Effective date: As of the date of issuance and shall be implemented
prior to reactor startup following the HNP, Unit 1, spring 2016,
refueling outage.
Amendment No.: 275. A publicly-available version is in ADAMS under
Accession No. ML15342A398; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-57: Amendment revised
the license and the Technical Specifications.
Date of initial notice in Federal Register: November 3, 2015 (80 FR
67802).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 29, 2016.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: April 29, 2015.
Brief description of amendment: The amendment revised the Cyber
Security Plan Implementation Milestone 8 completion date and the
physical protection license condition.
Date of issuance: January 28, 2016.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 214. A publicly-available version is in ADAMS under
Accession No. ML15328A059; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-30: The amendment
revised the Operating License.
Date of initial notice in Federal Register: July 7, 2015 (80 FR
38778).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 28, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 8th day of February 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-02916 Filed 2-12-16; 8:45 am]
BILLING CODE 7590-01-P