[Federal Register Volume 77, Number 113 (Tuesday, June 12, 2012)]
[Notices]
[Pages 35069-35079]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-13921]
[[Page 35069]]
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NUCLEAR REGULATORY COMMISSION
[NRC-2012-0131]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
Background
Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) is publishing this regular biweekly notice. The Act
requires the Commission publish notice of any amendments issued, or
proposed to be issued and grants the Commission the authority to issue
and make immediately effective any amendment to an operating license or
combined license, as applicable, upon a determination by the Commission
that such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 17, 2012 to May 30, 2012. The last
biweekly notice was published on May 29, 2012 (77 FR 31655).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and is publicly available, by
searching on http://www.regulations.gov under Docket ID NRC-2012-0131.
You may submit comments by the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0131. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2012-0131 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly available, by the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0131.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2012-0131 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 in comment submissions that you do not want to be publicly
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not 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 in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not edit comment submissions to remove such
information before making the comment submissions available to the
public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Title 10 of the Code of Federal
Regulations (10 CFR) 50.92, this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination; any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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
[[Page 35070]]
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) should
consult a current copy of 10 CFR 2.309, which is available at the NRC's
PDR, located at One White Flint North, Room O1-F21, 11555 Rockville
Pike (first floor), Rockville, Maryland 20852. The NRC regulations are
accessible electronically from the NRC Library on the NRC's Web site at
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or a presiding officer designated by the Commission or
by the Chief Administrative Judge of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the Chief Administrative Judge of the Atomic Safety and
Licensing Board will issue a notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through 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 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
[[Page 35071]]
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[email protected], or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information. With
respect to copyrighted works, except for limited excerpts that serve
the purpose of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: March 8, 2012.
Description of amendment request: The amendments would eliminate
the use of the term CORE ALTERATIONS throughout the Technical
Specifications (TSs). The proposed amendment incorporates changes
reflected in Technical Specification Task Force (TSTF) Change Traveler
TSTF-471-A, Revision 1, ``Eliminate use of term CORE ALTERATIONS in
ACTIONS and Notes.'' The U.S. Nuclear Regulatory Commission (NRC) staff
reviewed and approved TSTF-471 by letter dated December 7, 2006 (ADAMS
Accession No. ML062860320). The changes are consistent with NUREG-1432,
``Standard Technical Specifications--Combustion Engineering Plants,''
Revision 4 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML12102A165).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the use of the defined term CORE
ALTERATIONS from the Technical Specifications. CORE ALTERATIONS are
not an initiator of any accident previously evaluated except a fuel
handling accident. The revised Technical Specifications that protect
the initial conditions of a fuel handling accident also require the
suspension of movement of irradiated fuel assemblies. Suspending
movement of irradiated fuel assemblies protects the initial
condition of a fuel handling accident and, therefore, suspension of
CORE ALTERATIONS is not required. Suspension of CORE ALTERATIONS
does not provide mitigation of any accident previously evaluated.
Therefore, CORE ALTERATIONS do not affect the initiators of the
accidents previously evaluated and suspension of CORE ALTERATIONS
does not affect the mitigation of the accidents previously
evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical modification of the
plant (i.e., no new or different type of equipment will be
installed) or a significant 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. The proposed changes are 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 previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Only two accidents are postulated to occur during plant
conditions where CORE ALTERATIONS may be made: a fuel handling
accident and a boron dilution
[[Page 35072]]
accident. Suspending movement of irradiated fuel assemblies prevents
a fuel handling accident. Also requiring the suspension of CORE
ALTERATIONS is a redundant requirement to suspending movement of
irradiated fuel assemblies and does not increase the margin of
safety. CORE ALTERATIONS have no effect on a boron dilution
accident. Core components are not involved in the initiation or
mitigation of a boron dilution accident and the SHUTDOWN MARGIN
limit is based on assuming the worse-case configuration of the core
components.
Therefore, CORE ALTERATIONS have no effect on the margin of
safety related to a boron dilution accident.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three 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.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit 2, New London County, Connecticut
Date of amendment request: April 13, 2012.
Description of amendment request: The proposed amendment would
revise the Millstone Power Station, Unit 2 (MPS2) Technical
Specification (TS) requirements related to diesel fuel oil testing
consistent with NUREG-1432, Rev. 3.1, ``Standard Technical
Specifications, Combustion Engineering Plants,'' December 1, 1995, and
NRC approved Technical Specification Task Force (TSTF) TSTF-374,
``Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil,''
Revision 0.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
Criterion 1
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes modify the TS requirements related to
diesel fuel oil testing consistent with NRC approved TSTF-374,
``Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel
Oil,'' Revision 0. To adopt changes consistent with the content of
TSTF-374 for use in the custom TS of MPS2, the existing MPS2 diesel
fuel oil testing program will be modified. These changes replace the
criteria of ``Water and sediment < 0.05%'' with the criteria of ``A
clear and bright appearance with proper color or a water and
sediment content within limits'' and remove specific American
Society for Testing and Materials (ASTM) standard references from
TS.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems, and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences or any accident previously
evaluated.
Criterion 2
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are used to provide operational flexibility
regarding evolving industry standards while maintaining operational
conditions which are consistent with the design basis. Removing of
specific details from TS, since the details are already specified in
licensee-controlled documents, provides the flexibility needed to
maintain state-of-the-art technology in fuel oil sampling and
analysis methodology. The procedural details associated with the
involved specifications that are removed from TS and residing in
licensee-controlled documents are not required to be in the TS to
provide adequate protection of the public health and safety, since
the TS still retains the requirement for compliance with applicable
standards. 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 the provision, maintaining, or use of diesel fuel oil.
The requirements retained in the TS continue to require testing of
the diesel fuel oil to ensure the proper functioning of the DGs.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Criterion 3
Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: No.
The proposed changes are consistent with the content of TSTF-374
for use in the custom TS of MPS2. These changes remove specific ASTM
standard references and a preventive maintenance cleaning
requirement from TS since the references and requirements are
already specified in licensee-controlled documents. The proposed
changes provide the flexibility needed to improve fuel oil sampling
and analysis methodologies while maintaining sufficient controls to
ensure continued quality of the fuel oil. The margin of safety
provided to the DGs by these detailed fuel specifications is
unaffected by the proposed changes since there continue to be TS
requirements to ensure fuel oil is of the appropriate quality for
emergency DG use and DG operability is unaffected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: George Wilson.
Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-456 and STN
50-457, Braidwood Station, Units 1 and 2 (Braidwood), Will County,
Illinois, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1
and 2 (Byron), Ogle County, Illinois
Date of amendment request: March 20, 2012.
Description of amendment request: The proposed amendment would
modify Braidwood and Byron Technical Specifications to permanently
exclude portions of the steam generator (SG) tube below the top of the
SG tubesheet from periodic SG tube inspections and plugging or repair
for Braidwood, Unit 2 and for Byron, Unit 2. In addition, the proposed
amendment would revise TS 5.6.9 to remove reference to the previous
temporary alternate repair criteria and provide reporting requirements
specific to the permanent alternate repair criteria.
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
[[Page 35073]]
consequences of an accident previously evaluated?
Response: No.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
that alters the steam generator (SG) inspection and reporting
criteria does not have a detrimental impact on the integrity of any
plant structure, system, or component that initiates an analyzed
event. The proposed change will not alter the operation of, or
otherwise increase the failure probability of any plant equipment
that initiates an analyzed accident.
Of the various accidents previously evaluated, the proposed
changes only affect the steam generator tube rupture (SGTR),
postulated steam line break (SLB), feedwater line break (FLB),
locked rotor and control rod ejection accident evaluations. Loss-of-
coolant accident (LOCA) conditions cause a compressive axial load to
act on the tube. Therefore, since the LOCA tends to force the tube
into the tubesheet rather than pull it out, it is not a factor in
this amendment request. Another faulted load consideration is a safe
shutdown earthquake (SSE); however, the seismic analysis of Model D5
SGs has shown that axial loading of the tubes is negligible during
an SSE.
During the SGTR event, the required structural integrity margins
of the SG tubes and the tube-to-tubesheet joint over the H* distance
will be maintained. Tube rupture in tubes with cracks within the
tubesheet is precluded by the constraint provided by the presence of
the tubesheet and the tube-to-tubesheet joint. Tube burst cannot
occur within the thickness of the tubesheet. The tube-to-tubesheet
joint constraint results from the hydraulic expansion process,
thermal expansion mismatch between the tube and tubesheet, and from
the differential pressure between the primary and secondary side,
and tubesheet rotation. Based on this design, the structural margins
against burst, as discussed in draft Regulatory Guide (RG) 1.121,
``Bases for Plugging Degraded PWR Steam Generator Tubes,'' and TS
5.5.9, are maintained for both normal and postulated accident
conditions.
The proposed change has no impact on the structural or leakage
integrity of the portion of the tube outside of the tubesheet. The
proposed change maintains structural and leakage integrity of the SG
tubes consistent with the performance criteria of TS 5.5.9.
Therefore, the proposed change results in no significant increase in
the probability of the occurrence of a SGTR accident.
At normal operating pressures, leakage from tube degradation
below the proposed limited inspection depth is limited by the tube-
to-tubesheet crevice. Consequently, negligible normal operating
leakage is expected from degradation below the inspected depth
within the tubesheet region.
The consequences of an SGTR event are not affected by the
primary-to-secondary leakage flow during the event as primary-to-
secondary leakage flow through a postulated tube that has been
pulled out of the tubesheet is essentially equivalent to a severed
tube. Therefore, the proposed change does not result in a
significant increase in the consequences of a SGTR.
Primary-to-secondary leakage from tube degradation in the
tubesheet area during operating and accident conditions is
restricted due to contact of the tube with the tubesheet. The
leakage is modeled as flow through a porous medium through the use
of the Darcy equation. The leakage model is used to develop a
relationship between operational leakage and leakage at accident
conditions that is based on differential pressure across the
tubesheet and the viscosity of the fluid. A leak rate ratio was
developed to relate the leakage at operating conditions to leakage
at accident conditions. Since the fluid viscosity is based on fluid
temperature and it is shown that for the most limiting accident, the
fluid temperature does not exceed the normal operating temperature
and therefore the viscosity ratio is assumed to be 1.0. Therefore,
the leak rate ratio is a function of the ratio of the accident
differential pressure and the normal operating differential
pressure.
The leakage factor of 1.93 for Braidwood Station Unit 2 and
Byron Station Unit 2, for a postulated SLB/FLB, has been calculated
as shown in Table 9-7 of WCAP-17072-P, Revision 0. However, EGC
Braidwood Station Unit 2 and Byron Station Unit 2 will apply a
factor of 3.11 as determined by Westinghouse evaluation LTR-SGMP-09-
100 P-Attachment, Revision 1, to the normal operating leakage
associated with the tubesheet expansion region in the condition
monitoring (CM) and operational assessment (OA). The leakage factor
of 3.11 applies specifically to Byron Unit 2 and Braidwood Unit 2,
both hot and cold legs, in Table RAI24-2 of LTR-SGMP-09-100 P-
Attachment, Revision 1. Through application of the limited tubesheet
inspection scope, the existing operating leakage limit provides
assurance that excessive leakage (i.e., greater than accident
analysis assumptions) will not occur. The assumed accident induced
leak rate limit is 0.5 gallons per minute at room temperature
(gpmRT) for the faulted SG and 0.218 gpmRT for each of the unfaulted
SGs for accidents that assume a faulted SG. These accidents are the
SLB and the locked rotor with a stuck open PORV. The assumed
accident induced leak rate limit for accidents that do not assume a
faulted SG is 1.0 gpmRT for all SGs. These accidents are the locked
rotor and control rod ejection.
No leakage factor will be applied to the locked rotor or control
rod ejection transients due to their short duration, since the
calculated leak rate ratio is less than 1.0.
The TS 3.4.13 operational leak rate limit is 150 gallons per day
(gpd) (0.104 gpmRT) through any one SG. Consequently, there is
sufficient margin between accident leakage and allowable operational
leakage. The maximum accident leak rate ratio for the Model D5
design SGs is 1.93 as indicated in WCAP-17072-P, Revision 0, Table
9-7. However, EGC will use the more conservative value of 3.11
accident leak rate ratio for the most limiting SG model design
identified in Table RAI24-2 of LTRSGMP-09-100 P-Attachment Revision
1. This results in significant margin between the conservatively
estimated accident leakage and the allowable accident leakage (0.5
gpmRT).
For the CM assessment, the component of leakage from the prior
cycle from below the H* distance will be multiplied by a factor of
3.11 and added to the total leakage from any other source and
compared to the allowable accident induced leakage limit. For the
OA, the difference in the leakage between the allowable leakage and
the accident induced leakage from sources other than the tubesheet
expansion region will be divided by 3.11 and compared to the
observed operational leakage.
Based on the above, the performance criteria of NEI-97-06,
Revision 3, and draft RG 1.121 continue to be met and the proposed
change does not involve a significant increase in the probability or
consequences of the applicable accidents 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 introduce any changes or mechanisms
that create the possibility of a new or different kind of accident.
Tube bundle integrity is expected to be maintained for all plant
conditions upon implementation of the permanent alternate repair
criteria. The proposed change does not introduce any new equipment
or any change to existing equipment. No new effects on existing
equipment are created nor are any new malfunctions introduced.
Therefore, based on the above evaluation, the proposed 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 change defines the safety significant portion of
the SG tube that must be inspected and repaired. WCAP-17072-P,
Revision 0, as modified by WCAP-17330-P, Revision 1, identifies the
specific inspection depth below which any type tube degradation has
no impact on the performance criteria in NEI 97-06, Revision 3,
`Steam Generator Program Guidelines.''
The proposed change that alters the SG inspection and reporting
criteria maintains the required structural margins of the SG tubes
for both normal and accident conditions. NEI 97-06, and draft RG
1.121 are used as the bases in the development of the limited
tubesheet inspection depth methodology for determining that SG tube
integrity considerations are maintained within acceptable limits.
Draft RG 1.121 describes a method acceptable to the NRC for meeting
General Design Criteria (GDC) 14, ``Reactor Coolant Pressure
Boundary,'' GDC 15, ``Reactor Coolant System Design,'' GDC 31,
``Fracture Prevention of Reactor Coolant Pressure Boundary,'' and
GDC 32, ``Inspection of Reactor Coolant Pressure Boundary,'' by
reducing the probability and consequences of a SGTR. Draft RG 1.121
concludes that by determining the limiting safe conditions for tube
wall degradation, the probability and consequences of a SGTR are
reduced. This draft RG uses safety factors on loads for tube burst
that are consistent with
[[Page 35074]]
the requirements of Section III of the American Society of
Mechanical Engineers (ASME) Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, WCAP-17072-P, Revision 0, as
modified by WCAP-17330-P, Revision 1, defines a length of
degradation-free expanded tubing that provides the necessary
resistance to tube pullout due to the pressure induced forces, with
applicable safety factors applied. Application of the limited hot
and cold leg tubesheet inspection criteria will preclude
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage as described in
WCAP-17072-P, Revision 0, as modified by LTR-SGMP-09-100 P-
Attachment\ shows that significant margin exists between an
acceptable level of leakage during normal operating conditions that
ensures meeting the SLB accident-induced leakage assumption and the
TS leakage limit of 150 gpd.
Based on the above, it is concluded that the proposed changes do
not result in any reduction in a margin of safety.
Based on the above, EGC concludes that the proposed change
presents no 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: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road Warrenville, IL 60555.
NRC Branch Chief: Jacob I. Zimmerman.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County,
Texas
Date of amendment request: March 28, 2012.
Brief description of amendment: The amendment would revise
Technical Specification (TS) 5.5.9, ``Unit 1 Model D76 and Unit 2 Model
D5 Steam Generator (SG) Program,'' to permanently exclude portions of
the Comanche Peak Nuclear Power Plant (CPNPP), Unit 2, Model D5 SG
tubes below the top of the SG tubesheet from periodic SG tube
inspections. In addition, this amendment would revise TS 5.6.9, ``Unit
1 Model D76 and Unit 2 Model D5 Steam Generator Tube Inspection
Report,'' to provide permanent reporting requirements specific to
CPNPP, Unit 2, that have previously been established on a one-cycle
basis.
The proposed amendment constitutes a redefinition of the SG tube
primary-to-secondary pressure boundary and defines the safety
significant portion of the tube that must be inspected or plugged. Tube
flaws detected below the safety significant portion of the tube are not
required to be plugged. Allowing flaws in the non-safety significant
portion of the tube to remain in service minimizes unnecessary tube
plugging and maintains the safety margin of the steam generators to
perform the safety function to maintain the reactor coolant pressure
boundary, maintain reactor coolant flow, and maintain primary to
secondary heat transfer.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Of the accidents previously evaluated, the limiting transients
with consideration to the proposed change to the SG tube inspection
and repair criteria are the steam generator tube rupture (SGTR)
event, the steam line break (SLB), and the feed line break (FLB)
postulated accidents.
The required structural integrity margins of the SG tubes and
the tube-to-tubesheet joint over the H* distance will be maintained.
Tube rupture in tubes with cracks within the tubesheet is precluded
by the constraint provided by the presence of the tubesheet and the
tube-to-tubesheet joint. Tube burst cannot occur within the
thickness of the tubesheet. The tube-to-tubesheet joint constraint
results from the hydraulic expansion process, thermal expansion
mismatch between the tube and tubesheet, differential pressure
between the primary and secondary side, and tubesheet rotation.
Based on this design, the structural margins against burst, as
discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging
Degraded PWR [Pressurized Water Reactor] Steam Generator Tubes,''
[(Agencywide Documents Access and Management System (ADAMS)
Accession No. ML082120667)] and TS 5.5.9 are maintained for both
normal and postulated accident conditions.
The proposed change has no impact on the structural or leakage
integrity of the portion of the tube outside of the tubesheet. The
proposed change maintains structural and leakage integrity of the SG
tubes consistent with the performance criteria in TS 5.5.9.
Therefore, the proposed change results in no significant increase in
the probability of the occurrence of [an] SGTR accident.
At normal operating pressures, leakage from tube degradation
below the proposed limited inspection depth is limited by the tube-
to-tubesheet crevice. Consequently, negligible normal operating
leakage is expected from degradation below the inspected depth
within the tubesheet region. The consequences of an SGTR event are
not affected by the primary-to-secondary leakage flow during the
event as primary-to-secondary leakage flow through a postulated tube
that has been pulled out of the tubesheet is essentially equivalent
to a severed tube. Therefore, the proposed change does not result in
a significant increase in the consequences of [an] SGTR.
The probability of [an] SLB is unaffected by the potential
failure of a steam generator tube as the failure of tube is not an
initiator for [an] SLB event.
The leakage factor of 3.16 for CPNPP Unit 2, for a postulated
SLB/FLB, has been calculated as described in Westinghouse [Electric
Company, LLC] Letter LTR-SGMP-09-100 [N]P--Attachment, ``Response to
NRC Request for Additional Information on H*; Model F and Model D5
Steam Generators,'' dated August 12, 2009 [(ADAMS Accession No.
ML101730391)], and is shown in Revised Table 9-7 of this same
document. Specifically, for the condition monitoring (CM)
assessment, the component of leakage from the prior cycle from below
the H* distance will be multiplied by a factor of 3.16 and added to
the total leakage from any other source and compared to the
allowable accident induced leakage limit. For the operational
assessment (OA), the difference in the leakage between the allowable
leakage and the accident induced leakage from sources other than the
tubesheet expansion region will be divided by 3.16 and compared to
the observed operational leakage. The accident-induced leak rate
limit for CPNPP Unit 2 is 1.0 gpm [gallons per minute]. The TS
operational leak rate limit through any one steam generator is 150
gpd [gallons per day] (0.1 gpm). Consequently, there is significant
margin between accident leakage and allowable operational leakage.
The SLB/FLB overall leakage factor is 3.16 resulting in significant
margin between the conservatively estimated accident induced leakage
and the allowable accident leakage.
No leakage factor was applied to the locked rotor or control rod
ejection transients due to their short duration.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
that alters the SG inspection and reporting criteria does not have a
detrimental impact on the integrity of any plant structure, system,
or component that initiates an analyzed event. The proposed change
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change that alters the steam generator inspection
and reporting criteria
[[Page 35075]]
does not introduce any new equipment, create new failure modes for
existing equipment, or create any new limiting single failures.
Plant operation will not be altered, and all safety functions will
continue to perform as previously assumed in accident analyses.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The proposed change that alters the steam generator inspection
and reporting criteria maintains the required structural margins of
the SG tubes for both normal and accident conditions. Nuclear Energy
Institute [(NEI) document NEI] 97-06, Rev. 3, ``Steam Generator
Program Guidelines,'' and NRC Regulatory Guide (RG) 1.121, ``Bases
for Plugging Degraded PWR Steam Generator Tubes,'' are used as the
bases in the development of the limited tubesheet inspection depth
methodology for determining that SG tube integrity considerations
are maintained within acceptable limits. RG 1.121 describes a method
acceptable to the NRC for meeting General Design Criteria (GDC) 14,
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant
system design,'' GDC 31, ``Fracture prevention of reactor coolant
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant
pressure boundary,'' by reducing the probability and consequences of
a SGTR. RG 1.121 concludes that by determining the limiting safe
conditions for tube wall degradation, the probability and
consequences of a SGTR are reduced. RG 1.121 uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the American Society of Mechanical Engineers (ASME)
Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, the H* Analysis documented in
Section 4.1 [of the application dated March 28, 2012] defines a
length of degradation-free expanded tubing that provides the
necessary resistance to tube pullout due to the pressure induced
forces, with applicable safety factors applied. Application of the
limited hot and cold leg tubesheet inspection criteria will preclude
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage provides for
large margins between calculated and actual leakage values in the
proposed limited tubesheet inspection depth criteria.
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1800 M Street NW., Washington, DC 20036.
NRC Branch Chief: Michael T. Markley.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of amendment request: September 29, 2011, as supplemented by
letter dated March 12, 2012.
Description of amendment request: The proposed amendment would
revise the DAEC Technical Specifications (TSs) by modifying existing
Surveillance Requirements (SRs) regarding various modes of operation of
the main steam safety/relief valves (SRVs). The proposed amendment
would modify the TS requirements for testing of the SRVs by replacing
the current requirement to manually actuate each SRV during plant
startup with a series of overlapping tests that demonstrate the
required functions of successive valve stages. Elimination of the
manual actuation requirement at low reactor pressure and steam flow
decreases the potential for SRV leakage and spurious SRV opening.
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 any accident previously
evaluated?
Response: No.
The proposed changes modify TS SR 3.4.3.2, SR 3.5.1.9, and SR
3.6.1.5.1 to provide an alternative means for testing the main steam
SRVs, ADS [Automatic Depressurization System] valves, and LLS [Low-
Low Set] relief valves. Accidents are initiated by the malfunction
of plant equipment, or the catastrophic failure of plant structures,
systems, or components. The performance of SRV testing is not a
precursor to any accident previously evaluated and does not change
the manner in which the valves are operated. The proposed testing
requirements will not contribute to the failure of the SRVs nor any
plant structure, system, or component. NextEra Energy Duane Arnold
has determined that the proposed change in testing methodology
provides an equivalent level of assurance that the SRVs are capable
of performing their intended safety functions. Thus, the proposed
changes do not affect the probability of an accident previously
evaluated.
The performance of SRV testing provides confidence that the
relief valves are capable of depressurizing the reactor pressure
vessel (RPV). This will protect the reactor vessel from
overpressurization and allow the combination of the Low Pressure
Coolant Injection and Core Spray Systems to inject into the RPV as
designed. The LLS relief logic causes two LLS relief valves to be
opened at a lower pressure than the relief mode pressure setpoints
and causes the LLS relief valves to stay open longer, such that
reopening of more than one valve is prevented on subsequent
actuations. Thus, the LLS relief function prevents excessive short
duration SRV cycling, which limits induced thrust loads on the SRV
discharge line for subsequent actuations of the relief valve. The
proposed changes do not affect any function related to the safety
mode of the dual function SRVs. The proposed changes involve the
manner in which the subject valves are tested, and have no effect on
the types or amounts of radiation released or the predicted offsite
doses in the event of an accident. The proposed testing requirements
are sufficient to provide confidence that these valves are capable
of performing their intended safety functions.
In addition, an inadvertent opening of an SRV is an analyzed
event in the DAEC UFSAR [Updated Final Safety Analysis Report]
(Section 15.1.7.2), as well as the assumption of a single SRV
failure to open on demand in other transients and accidents, as
appropriate (e.g., one ADS valve failure in the LOCA [loss-of-
coolant accident] analysis). Since the proposed testing requirements
do not alter the assumptions for any analyzed transient or accident,
the radiological consequences of any accident previously evaluated
are not increased.
Therefore, the change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed changes do not affect the assumed accident
performance of the main steam SRVs, nor any plant structure, system,
or component previously evaluated. The proposed changes do not
install any new equipment, and installed equipment is not being
operated in a new or different manner. The proposed change in test
methodology will ensure that the valves remain capable of performing
their safety functions due to meeting the testing requirements of
the American Society of Mechanical Engineers Boiler and Pressure
Vessel Code, with the exception of opening the valve following
installation or maintenance for which a relief request has been
submitted (Ref. 6.1 [of the September 29, 2011, application]),
proposing an acceptable alternative. No setpoints are being changed
which would alter the dynamic response of plant equipment.
Accordingly, no new failure modes are introduced.
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 the margin of safety?
[[Page 35076]]
Response: No.
Overpressure protection of the RCPB [reactor coolant pressure
boundary] is based on the SRVs' setpoints and total relief capacity.
The setpoints are verified at an offsite testing facility; this
requirement is not altered by the proposed change. The relief
capacity of each SRV is determined by the valve's geometry, which is
also not altered by the proposed test methods.
The proposed changes will allow testing of the valve actuation
electrical circuitry, including the solenoid, and mechanical
actuation components, without causing the SRV to open. The SRVs will
be manually actuated prior to installation in the plant. Therefore,
all modes of SRV operation will be tested prior to entering the mode
of operation requiring the valves to perform their safety functions.
The proposed changes do not affect the valve setpoint or the
operational criteria that cause the SRVs to open during plant
transients or accidents, either manually or automatically. There are
no changes proposed which alter the setpoints at which protective
actions are initiated, and there is no change to the operability
requirements for equipment assumed to operate for accident
mitigation.
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. Mitchell S. Ross, P. O. Box 14000 Juno
Beach, FL 33408-0420.
NRC Acting Branch Chief: Istvan Frankl.
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: April 6, 2012, and revised on April 12
and May 7, 2012.
Description of amendment request: The proposed changes would amend
Combined License Nos. NPF-91 and NPF-92 for Vogtle Electric Generating
Plant (VEGP) Units 3 and 4, respectively, in regard to the upper
tolerance on the Nuclear Island (NI) critical sections basemat
thickness as identified in the plant-specific Design Control Document
(DCD).
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.
As indicated in FSAR (plant-specific DCD) Subsection 3.8.5.5,
the design function of the basemat is to provide the interface
between the nuclear island structures and the supporting soil or
rock. The basemat transfers the load of nuclear island structures to
the supporting soil or rock. The basemat transmits seismic motions
from the supporting soil or rock to the nuclear island. The revision
of the basemat construction tolerance does not have an adverse
impact on the response of the basemat and nuclear island structures
to safe shutdown earthquake ground motions or loads due to
anticipated transients or postulated accident conditions. The
revision of the basemat construction tolerance does not impact the
support, design, or operation of mechanical and fluid systems. 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 normal operation or postulated accident
conditions. The plant response to previously evaluated accidents or
external events is not adversely affected, nor does the change
described create any new accident precursors.
Therefore, there is no significant increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is to increase the construction tolerance
for the basemat thickness. The revision of the basemat construction
tolerance does not change the design of the basemat or nuclear
island structures. The revision of the basemat construction
tolerance does not change the design function, support, design, or
operation of mechanical and fluid systems. The revision of the
basemat construction tolerance does not result in a new failure
mechanism for the basemat or new accident precursors. As a result,
the design function of the basemat is not adversely affected by the
proposed change.
Therefore, the proposed change will 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 revision in the basemat thickness construction tolerance
does not have an adverse impact on the strength of the basemat. The
increase in the basemat thickness construction tolerance does not
have an adverse impact on the seismic design spectra or the
structural analysis of the basemat or other nuclear island
structures. The revision in the basemat thickness construction
tolerance has no impact of the analysis of the nuclear island for
sliding or overturning. As a result, the design function of the
basemat is not adversely affected by 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: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Mark E. Tonacci.
Virginia Electric and Power Company, Docket No. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of amendment request: April 2, 2012.
Description of amendment request: The proposed amendment would
delete the Steam Generator Water Level Low Coincident with Steam Flow/
Feedwater Flow Mismatch Reactor Trip Function from the Technical
Specification (TS) Table 3.3.1-1 Item 15.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The initiating conditions and assumptions for accidents
described in the Updated Final Safety Analyses Report remain as
previously analyzed. The proposed change does not introduce a new
accident initiator nor does it introduce changes to any existing
accident initiators or scenarios described in the Updated Final
Safety Analyses Report. The Steam/Feedwater Flow Mismatch and Low
Steam Generator Water Level reactor trip is not credited for
accident mitigation in any accident analyses described in the
Updated Final Safety Analyses Report. The Steam/Feedwater Flow
Mismatch and Low Steam Generator Water Level trip was designed to
meet the control and protection systems interaction criteria of
IEEE-279. The Steam Generator Level Median Signal Selector (MSS)
prevents adverse control and protection system interaction such that
it replaces the need for the Steam/Feedwater Flow Mismatch and Low
Steam Generator Water Level reactor trip to satisfy the IEEE-279
requirements. As such, the affected control and protection systems
will continue to perform their required functions without adverse
interaction, and maintain the capability to shut down the reactor
when required on Low-Low Steam Generator water level. The ability to
mitigate a loss of heat
[[Page 35077]]
sink accident previously evaluated is unaffected. The frequency
categories of previously evaluated accidents are not changed.
Therefore, neither the probability of occurrence nor the
consequences of an accident previously evaluated is significantly
increased.
Criterion 2--Does the change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The substitution of the MSS for the Steam/Feedwater Flow
Mismatch and Low Steam Generator Water Level trip will not introduce
any new failure modes to the required protection functions. The MSS
only interacts with the feedwater control system. The Steam
Generator Water Level Low-Low protection function is not affected by
this change. Isolation devices upstream of the MSS circuitry ensure
that the Steam Generator Water Level Low-Low protection function is
not affected. The MSS is designed to reduce the frequency of system
failures through utilization of highly reliable components in a
configuration that relies on a minimum of additional equipment.
Components used in the MSS are of a quality consistent with low
failure rates and minimum maintenance requirements, and conform to
protection system requirements. Furthermore, the design provides the
capability for complete unit testing that provides unambiguous
determination of credible system failures. It is through these
features that the overall design of the MSS minimizes the occurrence
of undetected failures that may exist between test intervals.
Therefore, the possibility for a new or different kind of
accident from any accident previously evaluated is not created.
Criterion 3--Does this change involve a significant reduction in a
margin of safety?
The proposed amendment does not involve revisions to any safety
analysis limits or safety system settings that will adversely impact
plant safety. The proposed amendment does not alter the functional
capabilities assumed in a safety analysis for any system, structure,
or component important to the mitigation and control of design bases
accident conditions within the facility. Nor does this amendment
revise any parameters or operating restrictions that are assumptions
of a design basis accident. In addition, the proposed amendment does
not affect the ability of safety systems to ensure that the facility
can be placed and maintained in a shutdown condition for extended
periods of time.
The ability of the Steam Generator Water Level Low-Low reactor
trip function credited in the safety analysis to protect against a
sudden loss of heat sink event is not affected by the proposed
change: Since the Steam Generator Low-Low Level trip is credited
alone as providing complete protection for the accident transients
that result in low steam generator level, eliminating the Steam/
Feedwater Flow Mismatch and Low Steam Generator Water Level trip
will not change any safety analysis conclusion for any analyzed
accident described in the Updated Final Safety Analyses Report.
The MSS prevents adverse control and protection system
interaction such that it replaces the need for the Steam/Feedwater
Flow Mismatch and Low Steam Generator Water Level reactor trip and
satisfies the IEEE-279 requirements.
The proposed change improves the margin of safety since removal
of the Steam/Feedwater Flow Mismatch and Low Steam Generator Water
Level trip function decreases the potential for challenges to plant
safety systems, decreases the plant surveillance/maintenance
activity, and reduces plant complexity. These changes result in a
reduction in the potential for unnecessary plant transients.
The Technical Specifications continue to assure that the
applicable operating parameters and systems are maintained within
the design requirements and safety analysis assumptions. Therefore,
the elimination of this trip function will not result in a
significant reduction in the margin of safety as defined in the
Updated Final Safety Analyses Report or Technical Specifications.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Nancy L. Salgado.
Virginia Electric and Power Company, Docket No. 50-339, North Anna
Power Station, Unit 2, Louisa County, Virginia
Date of amendment request: May 11, 2012.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) 3.1.7, ``Rod Position
Indication'' to allow two demand position indicators in one or more
banks to be inoperable for up to 4 hours. This change is proposed as a
temporary change to the TS for the current operating cycle and is
proposed as a footnote to the current TS Limiting Condition for
Operation (LCO) Section 3.1.7, Condition D.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change provides a new Condition for two demand
position indicators inoperable in one or more banks. The
inoperability of two demand position indicators in one or more banks
does not directly affect any accident analysis or design basis
limits or cause any limit not to be met, because the demand position
indicator only provides the intended demand as determined by the rod
control system. The actual position of the control rods is
determined by use of the Rod Position Indications (RPIs) for each
control rod, or the movable incore detector system when the RPIs are
inoperable.
The inoperability of the demand position indicators does prevent
the comparison of the RPIs to the demand position indication for
verification of rod insertion and rod group alignment limits, which
is conducted as a periodic surveillance to maintain the reactor
within analyzed conditions. The use of a 4 hour Completion Time
limit provides a restriction that limits the time that reactor
operation can continue during this loss of the demand position
indication. Since the loss of the demand position indication does
not cause the rods to change position, hence the actual control rod
positions are expected to remain within required limits. Placing the
Rod Control System in a condition incapable of rod movement is a
positive control to prevent rod stepping while maintenance is being
performed.
The proposed change to allow two demand position indicators to
be inoperable in one or more banks does not affect the automatic or
manual shutdown capability of the reactor protection system and no
accident analyses are impacted by the proposed change. The
operability of the control rods is not affected by the inoperability
of the demand position indicators.
Therefore, neither the probability of occurrence nor the
consequences of an accident previously evaluated is significantly
increased.
Criterion 2--Does the change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change provides new requirements for two demand
position indicators inoperable in one or more banks. No new accident
initiators are introduced by the proposed requirements because the
allowed condition for inoperability of the demand position
indicators does not cause any new failure modes to be created that
can cause an accident. The proposed change does not affect the
reactor protection system or the reactor control system. The control
rods should remain within the required limits because the failure of
the demand position indicators does not cause the rods to change
position and the RPIs remain available in the affected banks to
verify the position of the control rods. In addition, the Rod
Control System is placed in a condition incapable of rod movement as
a positive control to prevent rod stepping while maintenance is
[[Page 35078]]
being performed. Hence, no new failure modes or accident sequences
are created that would cause a new or different kind of accident
from any accident previously evaluated.
Therefore, the possibility for a new or different kind of
accident from any accident previously evaluated is not created.
Criterion 3--Does this change involve a significant reduction in a
margin of safety?
The operability of the RPIs is required to determine control rod
positions and thereby ensure compliance with the control rod
alignment and insertion limits. The proposed change does not alter
the requirement to determine rod position, but provides a new
Condition for two demand position indicators inoperable in one or
more banks. The inoperability of two demand position indicators for
one or more banks results in the reduced ability to periodically
verify that RPIs are operable and within expected limits. The
condition does prevent the comparison of the RPIs to the demand
position indication for verification of rod insertion and rod group
alignment limits, which is conducted as periodic surveillance to
maintain the reactor within analyzed conditions. The loss of the
demand position indication does not cause the rods to change
position, hence the actual control rod positions are expected to
remain within required limits. The use of a 4 hour Completion Time
limit provides a restriction that limits the time that reactor
operation can continue during this loss of the demand position
indication. This ensures the condition is promptly corrected or the
reactor shutdown in accordance with the applicable Technical
Specifications action statements. Thus, the proposed change
maintains the operation of the reactor within the applicable margins
of safety because the inoperability will be corrected or the unit
will be shutdown prior to any significant reduction in the ability
to verify control rod position by the use analog RPIs.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Nancy L. Salgado.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: November 30, 2011.
Description of amendment request: The proposed amendment would
revise the Wolf Creek Generating Station Technical Specification (TS)
3.8.1, ``AC Sources--Operating,'' Surveillance Requirements related to
Diesel Generator test loads, voltage, and frequency. The proposed
changes will correct non-conservative Diesel Generator load values that
are currently under administrative controls.
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 diesel generators are required to be OPERABLE in the event
of a design basis accident coincident with a loss of offsite power
to mitigate the consequences of the accident. The diesel generators
are not accident initiators and therefore these changes do not
involve a significant increase in the probability of an accident
previously evaluated.
The accident analyses assume that at least one engineered safety
feature bus is provided with power either from the offsite circuits
or the diesel generators. The Technical Specification change
proposed in this license amendment request will continue to assure
that the diesel generators have the capacity and capability to
assume their maximum design basis accident loads. The proposed
change does not significantly change how the plant would mitigate an
accident previously evaluated.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed change does not
adversely affect the ability of structures, systems, and components
(SSC) to perform their intended safety function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of any accident previously
evaluated. Further, the proposed change does not increase the types
and amounts of radioactive effluent that may be released offsite,
nor significantly increase individual or cumulative occupational/
public radiation exposure.
Therefore, the proposed change does not represent a significant
increase 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 Technical Specification change does not involve a
change in the plant design, system operation, or the use of the
diesel generators. The proposed change requires the diesel
generators to be tested at increased loads which envelope the actual
power demand requirements for the diesel generators during design
basis conditions. These revised loads continue to demonstrate the
capability and capacity of the diesel generators to perform their
required functions. There are no new failure modes or mechanisms
created due to testing the diesel generators at the proposed test
loading. Testing of the emergency diesel generators at the proposed
test loadings does not involve any modification in the operational
limits or physical design of plant systems. There are no new
accident precursors generated due to the proposed test loadings.
Therefore, it is concluded that 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 Technical Specification change will continue to
demonstrate that the diesel generators meet the Technical
Specification definition of OPERABILITY, that is, the proposed tests
will demonstrate that the diesel generators will perform their
safety function and the necessary diesel generator attendant
instrumentation, controls, cooling, lubrication and other auxiliary
equipment required for the emergency diesel generators to perform
their safety function loads are also tested at these proposed
loadings. The proposed testing will also continue to demonstrate the
capability and capacity of the diesel generators to supply their
required loads for mitigating a design basis accident.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside the design basis.
Therefore, it is concluded that 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
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
[[Page 35079]]
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
available online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR's
Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: July 27, 2011, as supplemented by
letters dated September 16, 2011, and February 7, February 24, and
April 3, 2012.
Brief description of amendment: The amendment modified River Bend
Station's (RBS) Technical Specification (TS) 3.3.6.1, ``Primary
Containment and Drywell Isolation Instrumentation,'' to revise the
allowable value (AV) and related setpoints for the Main Steam Tunnel
Temperature functions 1.e, 3.f, and 4.h in TS Table 3.3.6.1-1. In
addition, the RBS's Emergency Action Levels will be revised to reflect
the changes to the AV and related setpoints in TS 3.3.6.1.
Date of issuance: May 30, 2012.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 174.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 7, 2012 (77 FR
6147). The supplemental letters dated September 16, 2011, and February
7, February 24, and April 3, 2012, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 30, 2012.
No significant hazards consideration comments received: No.
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: June 2, 2011, as supplemented
by letter dated November 10, 2011.
Brief description of amendments: The amendments modify Technical
Specification (TS) 3.1.2, ``Reactivity Anomalies,'' to change the
method used to perform the reactivity anomaly surveillance.
Specifically, the amendments allow performance of the surveillance
based on the difference between the monitored (i.e., actual) core
reactivity and the predicted core reactivity. The surveillance was
previously performed based on the difference between the monitored
control rod density and the predicted control rod density.
Date of issuance: May 25, 2012.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 284 and 287.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the License and TSs.
Date of initial notice in Federal Register: September 6, 2011 (76
FR 55129).
The letter dated November 10, 2011, provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the application beyond the scope
of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 25, 2012.
No significant hazards consideration comments received: No.
South Carolina Electric and Gas Company, Docket No. 50-395, Virgil C.
Summer, Nuclear Station (VCSNS), Unit 1, Jenkinsville, South Carolina
Date of application for amendment: August 11, 2011.
Brief description of amendment: This amendment revised the VCSNS
Technical Specification (TS) to allow an updating of the applicable
topical report in TS 6.9.1.11, ``Core Operating Limits Report'' to use
the three-dimensional Advanced Nodal Code neutronic model.
Date of Issuance: May 30, 2012.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No: 190.
Renewed Facility Operating License No. NPF-12: Amendment revises
the License and Technical Specifications.
Date of initial notice in Federal Register: October 11, 2011 (76 FR
62864).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 30, 2012.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 1st day of June, 2012.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2012-13921 Filed 6-11-12; 8:45 am]
BILLING CODE 7590-01-P