Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 11472-11492 [2015-04298]
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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. Background
In accordance with the Paperwork
Reduction Act of 1995 (44 U.S.C.
Chapter 35), the NRC is requesting
public comment on its intention to
request the OMB’s approval for the
information collection summarized
below.
1. The title of the information
collection: NRC Request for Information
Concerning Patient Release Practices.
2. OMB approval number: OMB
control number has not yet been
assigned to this proposed information
collection.
3. Type of submission: New.
4. The form number, if applicable: N/
A.
5. How often the collection is required
or requested: Once.
6. Who will be required or asked to
respond: Medical professional
organizations, physicians, patients,
patient advocacy groups, NRC and
Agreement State medical use licensees,
Agreement States, and other interested
individuals who use, receive, license or
have interest in the use of I–131 sodium
iodine (hereafter referred to as ‘‘I–131’’)
for the treatment of thyroid conditions.
7. The estimated number of annual
responses: A one-time collection
estimated to have 1,180 responses (620
medical community + 560 patients).
8. The estimated number of annual
respondents: 1,180 respondents (620
medical community + 560 patients).
9. The estimated number of hours
needed annually to comply with the
information collection requirement or
request: 457.5 hours (255 medical
community + 202.5 patients).
10. Abstract: The NRC is requesting a
one-time information collection that
will be solicited in a Federal Register
notice (FRN). The FRN will have
specific I–131 patient release questions
associated with: (1) Existing Web sites
that the responders believe provide
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access to clear and consistent patient
information about I–131 treatment
processes and procedures; (2)
information the responders believe
represent best practices used in making
informed decisions on releasing I–131
patients and stand alone or
supplemental voluntary patient/licensee
guidance acknowledgment forms, if
available; (3) an existing set of
guidelines that the responder developed
or received that provides instructions to
released patients; and (4) an existing
guidance brochure that the responder
believes would be acceptable for
nationwide distribution. The responses
will form the basis for patient release
guidance products developed in
response to the NRC’s April 28, 2014,
Staff Requirements—COMAMM–14–
0001/COMWDM–14–0001—
‘‘Background and Proposed Direction to
NRC Staff to Verify Assumptions Made
Concerning Patient Release Guidance.’’
The Commission, based on information
from patients and patient advocacy
groups, questioned the availability of
clear, consistent, patient friendly and
timely patient release information and
directed the staff to work with a wide
variety of stakeholders when developing
new guidance products. This
information collection effort was
developed to gain input from as many
stakeholders as possible. The NRC
solicitation in the Federal Register is to
obtain existing information from a
variety of stakeholders.
III. Specific Requests for Comments
The NRC is seeking comments that
address the following questions:
1. Is the proposed collection of
information necessary for the NRC to
properly perform its functions? Does the
information have practical utility?
2. Is the estimate of the burden of the
information collection accurate?
3. Is there a way to enhance the
quality, utility, and clarity of the
information to be collected?
4. How can the burden of the
information collection on respondents
be minimized, including the use of
automated collection techniques or
other forms of information technology?
Dated at Rockville, Maryland, this 25th day
of February, 2015.
For the Nuclear Regulatory Commission.
Tremaine Donnell,
NRC Clearance Officer, Office of Information
Services.
[FR Doc. 2015–04318 Filed 3–2–15; 8:45 am]
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NUCLEAR REGULATORY
COMMISSION
[NRC–2015–0041]
Biweekly Notice; Applications and
Amendments to Facility Operating
Licenses and Combined Licenses
Involving No Significant Hazards
Considerations
Nuclear Regulatory
Commission.
ACTION: Biweekly notice.
AGENCY:
Pursuant to Section 189a. (2)
of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear
Regulatory Commission (NRC) is
publishing this regular biweekly notice.
The Act requires the Commission to
publish notice of any amendments
issued, or proposed to be issued and
grants the Commission the authority to
issue and make immediately effective
any amendment to an operating license
or combined license, as applicable,
upon a determination by the
Commission that such amendment
involves no significant hazards
consideration, notwithstanding the
pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all
notices of amendments issued, or
proposed to be issued from February 5,
2015 to February 18, 2015. The last
biweekly notice was published on
February 17, 2015.
DATES: Comments must be filed by April
2, 2015. A request for a hearing must be
filed by May 4, 2015.
ADDRESSES: You may submit comments
by any of the following methods (unless
this document describes a different
method for submitting comments on a
specific subject):
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2015–0041. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov.
• 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:
Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington DC
20555–0001; telephone: 301–415–5411,
email: Shirley.Rohrer@nrc.gov.
SUMMARY:
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II. Notice of Consideration of Issuance
of Amendments to Facility Operating
Licenses and Combined Licenses and
Proposed No Significant Hazards
Consideration Determination
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and
Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC–2015–
0041 when contacting the NRC about
the availability of information for this
action. You may obtain publiclyavailable information related to this
action by any of the following methods:
• Federal rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2015–0041.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publiclyavailable documents online in the
ADAMS Public Documents collection at
https://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 pdr.resource@nrc.gov. The
ADAMS accession number for each
document referenced (if it is available in
ADAMS) is provided the first time that
it is mentioned in the SUPPLEMENTARY
INFORMATION section.
• NRC’s PDR: You may examine and
purchase copies of public documents at
the NRC’s PDR, Room O1–F21, One
White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
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B. Submitting Comments
Please include Docket ID NRC–2015–
0041, 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 will post all comment
submissions at https://
www.regulations.gov as well as enter the
comment submissions into ADAMS.
The NRC does not routinely edit
comment submissions to remove
identifying or contact information.
If you are requesting or aggregating
comments from other persons for
submission to the NRC, then you should
inform those persons not to include
identifying or contact information that
they do not want to be publicly
disclosed in their comment submission.
Your request should state that the NRC
does not routinely edit comment
submissions to remove such information
before making the comment
submissions available to the public or
entering the comment into ADAMS.
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The Commission has made a
proposed determination that the
following amendment requests involve
no significant hazards consideration.
Under the Commission’s regulations in
§ 50.92 of Title 10 of the Code of Federal
Regulations (10 CFR), this means that
operation of the facility in accordance
with the proposed amendment would
not (1) involve a significant increase in
the probability or consequences of an
accident previously evaluated, or (2)
create the possibility of a new or
different kind of accident from any
accident previously evaluated; or (3)
involve a significant reduction in a
margin of safety. The basis for this
proposed determination for each
amendment request is shown below.
The Commission is seeking public
comments on this proposed
determination. Any comments received
within 30 days after the date of
publication of this notice will be
considered in making any final
determination.
Normally, the Commission will not
issue the amendment until the
expiration of 60 days after the date of
publication of this notice. The
Commission may issue the license
amendment before expiration of the 60day 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
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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 https://
www.nrc.gov/reading-rm/doccollections/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
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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 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.
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 EFiling process requires participants to
submit and serve all adjudicatory
documents over the internet, or in some
cases to mail copies on electronic
storage media. Participants may not
submit paper copies of their filings
unless they seek an exemption in
accordance with the procedures
described below.
To comply with the procedural
requirements of E-Filing, at least ten 10
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days prior to the filing deadline, the
participant should contact the Office of
the Secretary by email at
hearing.docket@nrc.gov, 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 NRCissued 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 https://
www.nrc.gov/site-help/e-submittals/
getting-started.html. System
requirements for accessing the ESubmittal server are detailed in the
NRC’s ‘‘Guidance for Electronic
Submission,’’ which is available on the
agency’s public Web site at https://
www.nrc.gov/site-help/esubmittals.html. Participants may
attempt to use other software not listed
on the Web site, but should note that the
NRC’s E-Filing system does not support
unlisted software, and the NRC Meta
System Help Desk will not be able to
offer assistance in using unlisted
software.
If a participant is electronically
submitting a document to the NRC in
accordance with the E-Filing rule, the
participant must file the document
using the NRC’s online, Web-based
submission form. In order to serve
documents through the Electronic
Information Exchange System, users
will be required to install a Web
browser plug-in from the NRC’s Web
site. Further information on the Webbased submission form, including the
installation of the Web browser plug-in,
is available on the NRC’s public Web
site at https://www.nrc.gov/site-help/esubmittals.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 https://www.nrc.gov/site-help/esubmittals.html. A filing is considered
complete at the time the documents are
submitted through the NRC’s E-Filing
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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 https://
www.nrc.gov/site-help/esubmittals.html, by email to
MSHD.Resource@nrc.gov, or by a tollfree 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
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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 https://
ehd1.nrc.gov/ehd/, unless excluded
pursuant to an order of the Commission,
or the presiding officer. Participants are
requested not to include personal
privacy information, such as social
security numbers, home addresses, or
home phone numbers in their filings,
unless an NRC regulation or other law
requires submission of such
information. However, a request to
intervene will require including
information on local residence in order
to demonstrate a proximity assertion of
interest in the proceeding. With respect
to copyrighted works, except for limited
excerpts that serve the purpose of the
adjudicatory filings and would
constitute a Fair Use application,
participants are requested not to include
copyrighted materials in their
submission.
Petitions for leave to intervene must
be filed no later than 60 days from the
date of publication of this notice.
Requests for hearing, petitions for leave
to intervene, and motions for leave to
file new or amended contentions that
are filed after the 60-day deadline will
not be entertained absent a
determination by the presiding officer
that the filing demonstrates good cause
by satisfying the three factors in 10 CFR
2.309(c)(1)(i)–(iii).
For further details with respect to
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 No. 2
(ANO–2), Pope County, Arkansas
Date of amendment request: February
6, 2015. A publicly-available version is
in ADAMS under Accession No.
ML15041A068.
Description of amendment request:
The amendment would revise a Note to
Technical Specification (TS)
Surveillance Requirement (SR) 4.1.3.1.2
to exclude Control Element Assembly
(CEA) 18 from being exercised per the
SR for the remainder of Cycle 24 due to
a degrading upper gripper coil.
Basis for proposed no significant
hazards consideration determination:
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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.
One function of the CEAs is to provide a
means of rapid negative reactivity addition
into the core. This occurs upon receipt of a
signal from the Reactor Protection System.
This function will continue to be
accomplished with the approval of the
proposed change. Typically, once per 92 days
each CEA is moved at least five inches to
ensure the CEA is free to move. CEA 18
remains trippable (free to move) as illustrated
by the last performance of SR 4.1.3.1.2 in
January 2015. However, due to abnormally
high coil voltage and current measured on
the CEA 18 Upper Gripper Coil (UGC), future
exercising of the CEA could result in the CEA
inadvertently inserting into the core, if the
UGC were to fail during the exercise test. The
mis-operation of a CEA, which includes a
CEA drop event, is an abnormal occurrence
and has been previously evaluated as part of
the ANO–2 accident analysis. Inadvertent
CEA insertion will result in a reactivity
transient and power reduction, and could
lead to a reactor shutdown if the CEA is
deemed to be unrecoverable. The proposed
change would minimize the potential for
inadvertent insertion of CEA 18 into the core
by maintaining the CEA in place using the
Lower Gripper Coil (LGC), which is operating
normally. The proposed change will not
affect the CEAs ability to insert fully into the
core upon receipt of a reactor trip signal.
No modifications are proposed to the
Reactor Protection System or associated
Control Element Drive Mechanism Control
System logic with regard to the ability of CEA
18 to remain available for immediate
insertion. The accident mitigation features of
the plant are not affected by the proposed
amendment. Because CEA 18 remains
trippable, no additional reactivity
considerations need to be taken into
consideration. Nevertheless, Entergy has
evaluated the reactivity consequences
associated with failure of CEA 18 to insert
upon a reactor trip in accordance with TS
requirements for Shutdown Margin (SDM)
and has determined that SDM requirements
would be met should such an event occur at
any time during the remainder of Cycle 24
operation.
Therefore, this change does not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
CEA 18 remains trippable. The proposed
change will not introduce any new design
changes or systems that can prevent the CEA
from [performing] its specified safety
function. As discussed previously, CEA mis-
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operation has been previously evaluated in
the ANO–2 accident analysis. Furthermore,
SDM has been shown to remain within limits
should an event occur at any time during the
remainder of operating Cycle 24 such that
CEA 18 fails to insert into the core upon
receipt of a reactor trip signal.
Therefore, 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.
SR 4.1.3.1.2 is intended to verify CEAs are
free to move (i.e., not mechanically bound).
The physical and electrical design of the
CEAs, and past operating experience,
provides high confidence that CEAs remain
trippable whether or not exercised during
each SR interval. Eliminating further
exercising of CEA 18 for the remainder of
Cycle 24 operation does not directly relate to
the potential for CEA binding to occur. No
mechanical binding has been previously
experienced at ANO–2. CEA 18 is contained
within a Shutdown CEA Group and is not
used for reactivity control during power
maneuvers (the CEA must remain fully
withdrawn at all times when the reactor is
critical). In addition, Entergy has concluded
that required SDM will be maintained should
CEA 18 fail to insert following a reactor trip
at any point during the remainder of Cycle
24 operation.
Therefore, this 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 Acting Branch Chief: Eric R.
Oesterle.
Entergy Operations, Inc., Docket No. 50–
382, Waterford Steam Electric Station,
Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October
1, 2014, as supplemented by letter dated
February 2, 2015. Publicly-available
versions are in ADAMS under
Accession Nos. ML14275A374 and
ML15033A482.
Description of amendment request:
The amendment would relocate
Technical Specifications 3.9.6, ‘‘Refuel
Machine,’’ and 3.9.7, ‘‘Crane Travel,’’ to
the Technical Requirements Manual.
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
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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.
This proposed change relocates
Technical Specifications (TS) 3.9.6
(Refuel Machine) and TS 3.9.7 (Crane
Travel) to the Waterford 3 Technical
Requirements Manual (TRM). This is
consistent with the requirements of [10
CFR 50.36(c)(2)(ii)] and aligns with
NUREG–1432 (Combustion Engineering
Standard Technical Specifications).
The applicable TS 3.9.6 and TS 3.9.7
design basis accident is the Fuel
Handling Accident (FHA) described in
the Updated Final Safety Analysis
Report (UFSAR) Section 15.7.3.4. The
limiting FHA results in all the fuel pins
in the dropped and impacted fuel
assemblies failing (472 pins or 236 per
assembly). The analysis assumes that a
fuel assembly is dropped as an initial
condition and no equipment or
intervention can prevent the initiating
condition. The proposed change was
evaluated against [10 CFR
50.36(c)(2)(ii)] criteria and shows no
impact to the lowest functional
capability or performance levels of
equipment required for safe operation of
the facility because the TS 3.9.6 and TS
3.9.7 requirements do not prevent the
accident conditions from occurring and
do not limit the severity of the accident.
Since, the dropped fuel assembly and
the impacted fuel assembly are both
already failed in the design basis
accident scenario, this change could not
result in a significant increase in the
accident consequences. The TS 3.9.6
and TS 3.9.7 equipment are not required
to respond, mitigate, or terminate any
design basis accident, thus this change
will not adversely impact the likelihood
or probability of a design basis accident.
The TS 3.9.6 and TS 3.9.7
requirements do not prevent the
accident conditions from occurring and
do not limit the severity of the accident.
Therefore the TS 3.9.6 and TS 3.9.7
relocation to the TRM would not cause
a significant increase in the accident
probability or accident consequences.
2. Does the proposed change create
the possibility of a new or different kind
of accident from any accident
previously evaluated?
Response: No.
This proposed change relocates TS
3.9.6 (Refuel Machine) and TS 3.9.7
(Crane Travel) to the Waterford 3 TRM.
In general, Technical Specifications are
based upon the accident analyses. The
accident analyses assumptions and
initial conditions must be protected by
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the Technical Specifications. This is a
requirement as outlined in [10 CFR
50.36].
[10 CFR 50.36(b)] states the technical
specifications will be derived from the
analyses and evaluation included in the
safety analysis report.
[10 CFR 50.36(c)(2)(i)] states that
[‘‘]the limiting conditions for operation
are the lowest functional capability or
performance levels of equipment
required for safe operation of the
facility[. . . .’’] [10 CFR 50.36(c)(2)(ii)]
provides the four criteria in which any
one met requires a limiting condition for
operation. The proposed change
demonstrated that the [10 CFR
50.36(c)(2)(ii)] criteria were not met and
the relocation to the TRM is allowable.
By not meeting the [10 CFR
50.36(c)(2)(ii)] criteria for inclusion into
the TS means that TS 3.9.6 and TS 3.9.7
do not impact the accident analyses
previously evaluated and would not
create the possibility of a new or
different kind of accident.
Specifically, TS 3.9.6 and TS 3.9.7
equipment are not instrumentation used
to detect, and indicate in the control
room, a significant abnormal
degradation of the reactor coolant
pressure boundary (Criterion 1). TS
3.9.6 and TS 3.9.7 do not contain a
process variable, design feature, or
operating restriction that is an initial
condition of a Design Basis Accident or
Transient analysis that either assumes
the failure of or presents a challenge to
the integrity of a fission product barrier
(Criterion 2). TS 3.9.6 and TS 3.9.7 does
not contain a structure, system, or
component that is part of the primary
success path and which functions or
actuates to mitigate a Design Basis
Accident or Transient that either
assumes the failure of or presents a
challenge to the integrity of a fission
product barrier (Criterion 3). Lastly, TS
3.9.6 and TS 3.9.7 do not contain a
structure, system, or component which
operating experience or probabilistic
safety assessment has shown to be
significant to public health and safety
(Criterion 4).
TS 3.9.6 and 3.9.7 are not required to
meet the lowest functional capability or
performance levels of equipment
required for safe operation of the
facility.
Therefore, the accident analyses are
not impacted and the possibility of a
new or different kind of accident from
any accident previously evaluated has
not changed.
3. Does the proposed change involve
a significant reduction in a margin of
safety?
Response: No.
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The proposed TS 3.9.6 (Refuel
Machine) and TS 3.9.7 (Crane Travel)
relocation to the Waterford 3 TRM is
administrative in nature because all
requirements will be relocated. Any
changes after being relocated to the
Waterford 3 TRM will require that the
[10 CFR 50.59] process be entered
ensuring the public health and safety is
maintained. By using the [10 CFR 50.59]
process for future changes, the
regulatory requirements ensure that no
significant reduction in the margin of
safety occurs.
In addition, the TS 3.9.6 and TS 3.9.7
requirements do not prevent the design
basis accident conditions from
occurring and do not limit the severity
of the accident. Thus, TS 3.9.6 and TS
3.9.7 relocation will not adversely
impact the accident analyses and will
not cause a significant reduction in the
margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Joseph A.
Aluise, Associate General Council—
Nuclear, Entergy Services, Inc., 639
Loyola Avenue, New Orleans, Louisiana
70113.
NRC Branch Chief: Meena K. Khanna.
Exelon Generation Company, LLC
(EGC), Docket No. 50–410, Nine Mile
Point Nuclear Station, Unit 2 (NMP2),
Oswego County, New York
Date of amendment request:
November 17, 2014. A publicly
available version is in ADAMS under
Accession No. ML14321A744.
Description of amendment request:
The proposed amendment would revise
the NMP2 Technical Specification (TS)
Allowable Value for the Main Steam
Line Tunnel Lead Enclosure
Temperature-High instrumentation from
an ambient temperature dependent
(variable setpoint) to ambient
temperature independent (constant
Allowable Value). The changes would
delete Surveillance Requirement (SR)
3.3.6.1.2 and revise the Allowable Value
for Function 1.g on Table 3.3.6.1–1,
‘‘Primary Containment Isolation
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:
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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 involve
a significant increase in the probability
or consequences of an accident
previously evaluated because the
performance of any equipment credited
in the radiological consequences of an
accident is not affected by the change in
the leak detection capability.
The Main Steam Line Tunnel Lead
Enclosure Temperature—High is
provided to detect a steam leak in the
lead enclosure and provides diversity to
the high flow instrumentation. This
function provides a mitigating action for
a steam leak in the Main Steam Line
Tunnel Lead Enclosure, which could
lead to a pipe break. This function does
not affect any accident precursors, and
the proposed changes do not affect the
leak detection capability. Additionally,
the proposed changes do not degrade
the performance of or increase the
challenges to any safety systems
assumed to function in the accident
analysis.
Therefore, the proposed changes do
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 do not create
the possibility of a new or different kind
of accident from any accident
previously evaluated because the
proposed changes do not add or remove
equipment and do not physically alter
the isolation instrumentation. In
addition, the Main Steam Line Tunnel
Lead Enclosure LDS [Leak Detection
System] is not utilized in a different
manner. The proposed changes do not
introduce any new accident initiators
and new failure modes, nor do they
reduce or adversely affect the
capabilities of any plant structure,
system, or component to perform their
safety function. The Main Steam Line
Tunnel Lead Enclosure LDS will
continue to be operated in the same
manner.
Therefore, the proposed changes do
not create the possibility of a new or
different kind of accident from any
accident previously evaluated.
3. Does the proposed amendment
involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not involve
a significant reduction in a margin of
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safety because the changes eliminate the
temperature setpoint dependency on
lead enclosure temperature while
maintaining the existing upper AV
[Allowable Value] = 175.6 °F, that was
previously evaluated and approved.
There is no adverse impact on the
existing equipment capability as well as
associated structures. The increase in
the steam leak rate and associated crack
size continues to be well below the leak
rate associated with critical crack size
that leads to pipe break. The proposed
changes continue to provide the same
level of protection against a main steam
line break as the existing setpoint
values.
Therefore, the proposed changes do
not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: J. Bradley
Fewell, Senior Vice President,
Regulatory Affairs, Nuclear, and General
Counsel, Exelon Generation Company,
LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Branch Chief: Benjamin G.
Beasley.
Florida Power and Light Company, et al.
(FPL), Docket Nos. 50–335 and 50–389,
St. Lucie Plant, Unit Nos. 1 and 2, St.
Lucie County, Florida
Date of amendment request: February
20, 2014, as supplemented by letters
dated December 11, 2014, January 13
and January 28, 2015. Publicly-available
in ADAMS under Accession Nos.
ML14070A087, ML14349A333,
ML15029A497 and ML15042A122.
Description of amendment request:
The NRC staff has previously made a
proposed determination that the
amendment request dated February 20,
2014, involves no significant hazards
consideration (see 79 FR 42550, July 22,
2014). Subsequently, by letter dated
January 28, 2015, the licensee provided
additional information that expanded
the scope of the amendment request as
originally noticed. Accordingly, this
notice supersedes the previous notice in
its entirety.
The amendment would revise the
Technical Specifications (TSs) by
relocating specific surveillance
frequency requirements to a licenseecontrolled program with
implementation of Nuclear Energy
Institute (NEI) 04–10 (Revision 1),
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11477
‘‘Risk-Informed Technical
Specifications Initiative 5b, RiskInformed Method for Control of
Surveillance Frequencies’’ (ADAMS
Accession No. ML071360456). The
licensee stated that the NEI 04–10
methodology provides reasonable
acceptance guidelines and methods for
evaluating the risk increase of proposed
changes to surveillance frequencies,
consistent with Regulatory Guide 1.177,
‘‘An Approach for Plant-Specific, RiskInformed Decisionmaking: Technical
Specifications’’ (ADAMS Accession No.
ML003740176). The licensee stated that
the changes are consistent with NRCapproved Technical Specification Task
Force (TSTF) Improved Standard
Technical Specifications Change
Traveler TSTF–425, ‘‘Relocate
Surveillance Frequencies to Licensee
Control—RITSTF [Risk-Informed
Technical Specifications Task Force]
Initiative 5b,’’ Revision 3 (ADAMS
Accession No. ML090850642). The
Federal Register notice published on
July 6, 2009 (74 FR 31996), announced
the availability of TSTF–425, Revision
3. In the supplement dated January 28,
2015, the licensee requested (1)
additional surveillance frequencies be
relocated to the licensee-controlled
program, (2) editorial changes, (3)
administrative deviations from TSTF–
425, and (4) other changes resulting
from differences between the St. Lucie
Plant TSs and the TSs on which TSTF–
425 was based.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee 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 relocates the
specified frequencies for periodic
surveillance requirements to licensee control
under a new Surveillance Frequency Control
Program. Surveillance frequencies are not an
initiator to any accident previously
evaluated. As a result, the probability of any
accident previously evaluated is not
significantly increased. The systems and
components required by the technical
specifications for which the surveillance
frequencies are relocated are still required to
be operable, meet the acceptance criteria for
the surveillance requirements, and be
capable of performing any mitigation
function assumed in the accident analysis.
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
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probability or consequences of any 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.
No new or different accidents result from
utilizing the proposed change. The changes
do not involve a physical alteration of the
plant (i.e., no new or different type of
equipment will be installed) or a change in
the methods governing normal plant
operation. In addition, the changes do not
impose any new or different requirements.
The changes do not alter assumptions made
in the safety analysis assumptions and
current plant operating practice.
Therefore, the proposed changes do not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a
significant reduction in the margin of safety?
Response: No.
The design, operation, testing methods,
and acceptance criteria for systems,
structures, and components (SSCs), specified
in applicable codes and standards (or
alternatives approved for use by the NRC)
will continue to be met as described in the
plant licensing basis (including the final
safety analysis report and bases to TS), since
these are not affected by changes to the
surveillance frequencies. Similarly, there is
no impact to safety analysis acceptance
criteria as described in the plant licensing
basis. To evaluate a change in the relocated
surveillance frequency, FPL will perform a
probabilistic risk evaluation using the
guidance contained in NRC-approved NEI
04–10, Revision 1 in accordance with the TS
Surveillance Frequency Control Program. NEI
04–10, Revision 1, methodology provides
reasonable acceptance guidelines and
methods for evaluating the risk increase of
proposed changes to surveillance frequencies
consistent with Regulatory Guide (RG) 1.177.
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Therefore, the proposed changes do
not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: William S.
Blair, Managing Attorney—Nuclear,
Florida Power & Light Company, 700
Universe Boulevard, MS LAW/JB, Juno
Beach, FL 33408–0420.
NRC Branch Chief: Shana R. Helton.
Florida Power and Light Company,
Docket Nos. 50–250 and 50–251, Turkey
Point Nuclear Generating, Unit Nos. 3
and 4, Miami-Dade County, Florida
Date of amendment request:
November 13, 2014. A publiclyavailable version is in ADAMS under
Accession No. ML14337A013.
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Description of amendment request:
The amendment would revise Technical
Specification (TS) 3/4.5.2, ‘‘ECCS
[Emergency Core Cooling System]
Subsystems—Tavg [average temperature]
Greater Than or Equal to 350 °F [degrees
Fahrenheit],’’ to correct nonconservative TS requirements. The
licensee also requested editorial changes
to the TS.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee provided its analysis of the
issue of no significant hazards
consideration, which is presented as
follows:
1. Does the proposed change involve
a significant increase in the probability
or consequences of an accident
previously evaluated?
No. The proposed TS changes involve
TS 3.5.2 Action ‘a’, new TS 3.5.2 Action
‘h’, and the provision in SR
[Surveillance Requirement] 4.5.2.a to
address non-conservative TS
requirements. Editorial changes are also
proposed for consistency and clarity.
These changes do not affect any
precursors to any accident previously
evaluated and subsequently, will not
impact the probability or consequences
of an accident previously evaluated.
Furthermore, these changes do not
adversely affect mitigation equipment or
strategies.
Therefore, the proposed changes do
not involve a significant increase in the
probability or consequences of an
accident previously evaluated.
2. Does the proposed change create
the possibility of a new or different kind
of accident from any previously
evaluated?
No. The proposed TS changes involve
TS 3.5.2 Action ‘a’, new TS 3.5.2 Action
‘h’, and the provision in SR 4.5.2.a to
address non-conservative TS
requirements. Editorial changes are also
proposed for consistency and clarity.
The proposed changes provide better
assurance that the ECCS systems,
subsystems, and components are
properly aligned to support safe reactor
operation consistent with the licensing
basis requirements. The proposed
changes do not introduce new modes of
plant operation and do not involve
physical modifications to the plant (no
new or different type of equipment will
be installed). There are no changes in
the method by which any safety related
plant structure, system, or component
(SSC) performs its specified safety
function. As such, the plant conditions
for which the design basis accident
analyses were performed remain valid.
No new accident scenarios, transient
precursors, failure mechanisms, or
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limiting single failures will be
introduced as a result of the proposed
change. There will be no adverse effect
or challenges imposed on any SSC as a
result of the proposed change.
Therefore, the proposed changes do
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 the margin of
safety?
No. Margin of safety is related to
confidence in the ability of the fission
product barriers to perform their
accident mitigation functions. The
proposed TS changes involve TS 3.5.2
Action ‘a’, new TS 3.5.2 Action ‘h’, and
the provision in SR 4.5.2.a to address
non-conservative TS requirements.
Editorial changes are also proposed for
consistency and clarity. The proposed
changes provide better assurance that
the ECCS systems, subsystems, and
components are properly aligned to
support safe reactor operation consistent
with the licensing basis requirements.
The proposed changes do not physically
alter any SSC. There will be no effect on
those SSCs necessary to assure the
accomplishment of specified functions.
There will be no impact on the
overpower limit, departure from
nucleate boiling ratio (DNBR) limits,
loss of cooling accident peak cladding
temperature (LOCA PCT), or any other
margin of safety. The applicable
radiological dose consequence
acceptance criteria will continue to be
met. Therefore, the proposed changes do
not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: William S.
Blair, Managing Attorney—Nuclear,
Florida Power & Light Company, P.O.
Box 14000, Juno Beach, FL 33408–0420.
NRC Branch Chief: Shana R. Helton.
Indiana Michigan Power Company
(I&M), Docket Nos. 50–315 and 50–316,
Donald C. Cook Nuclear Plant, Units 1
and 2, Berrien County, Michigan
Date of amendment request: February
6, 2015. A publicly-available version is
in ADAMS under Accession No.
ML15041A069.
Description of amendment request:
The proposed amendments would
modify the technical specifications
requirements for unavailable barriers by
adding limiting condition for operation
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(LCO) 3.0.8. The changes are consistent
with the NRC approved Technical
Specification Task Force (TSTF)
Standard Technical Specification
change TSTF–427, ‘‘Allowance for NonTechnical Specification Barrier
Degradation on Supported System
OPERABILITY,’’ Revision 2.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has affirmed the applicability
of the model proposed no significant
hazards consideration published on
October 3, 2006 (71 FR 58444), ‘‘Notice
of Availability of the Model Safety
Evaluation.’’ The findings presented in
that evaluation are presented below:
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Criterion 1—The Proposed Change Does
Not Involve a Significant Increase in the
Probability or Consequences of an
Accident Previously Evaluated
The proposed change allows a delay
time for entering a supported system
technical specification (TS) when the
inoperability is due solely to an
unavailable barrier if risk is assessed
and managed. The postulated initiating
events which may require a functional
barrier are limited to those with low
frequencies of occurrence, and the
overall TS system safety function would
still be available for the majority of
anticipated challenges. Therefore, the
probability of an accident previously
evaluated is not significantly increased,
if at all. The consequences of an
accident while relying on the allowance
provided by proposed LCO 3.0.8 are no
different than the consequences of an
accident while relying on the TS
required actions in effect without the
allowance provided by proposed LCO
3.0.8. 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, this change does
not involve a significant increase in the
probability or consequences of an
accident previously evaluated.
Criterion 2—The Proposed Change Does
Not Create the Possibility of a New or
Different Kind of Accident From any
Previously Evaluated
The proposed change does not
involve a physical alteration of the plant
(no new or different type of equipment
will be installed). Allowing delay times
for entering supported system TS when
inoperability is due solely to an
unavailable barrier, if risk is assessed
and managed, will not introduce new
failure modes or effects and will not, in
the absence of other unrelated failures,
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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 will
further minimize possible concerns.
Thus, this change does not create the
possibility of a new or different kind of
accident from an accident previously
evaluated.
Criterion 3—The Proposed Change Does
Not Involve a Significant Reduction in
a Margin of Safety
The proposed change allows a delay
time for entering a supported system TS
when the inoperability is due solely to
an unavailable barrier, if risk is assessed
and managed. The postulated initiating
events which may require a functional
barrier are limited to those with low
frequencies of occurrence, and the
overall TS system safety function would
still be available for the majority of
anticipated challenges. The risk impact
of the proposed TS changes was
assessed following the three-tiered
approach recommended in RG
[Regulatory Guide] 1.177. A bounding
risk assessment was performed to justify
the proposed TS changes. This
application of LCO 3.0.8 is predicated
upon the licensee’s performance of a
risk assessment and the management of
plant risk. The net change to the margin
of safety is insignificant as indicated by
the anticipated low levels of associated
risk (ICCDP [incremental conditional
core damage probability] and ICLERP
[incremental large early release
probability]) as shown in Table 1 of
Section 3.1.1 in the Safety Evaluation.
Therefore, this change does not involve
a significant reduction in a margin of
safety.
The NRC staff has reviewed the analysis
and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to
determine that the amendment requests
involve no significant hazards consideration.
Attorney for licensee: Robert B. Haemer,
Senior Nuclear Counsel, One Cook Place,
Bridgman, Michigan 49106.
NRC Branch Chief: David L. Pelton.
PPL Susquehanna, LLC, Docket Nos. 50–387
and 50–388, Susquehanna Steam Electric
Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment requests: October 27,
2014. A publicly-available version is
available in ADAMS under Accession No.
ML14317A052.
Description of amendment requests: The
proposed amendments will modify the
Susquehanna technical specifications (TS).
Specifically, the proposed amendments will
modify the TS by relocating specific
surveillance frequencies to a licenseecontrolled program, the Surveillance
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Frequency Control Program (SFCP), with
implementation of Nuclear Energy Institute
(NEI) 04–10, ‘‘Risk-Informed Technical
Specifications Initiative 5b, Risk-Informed
Method for Control of Surveillance
Frequencies’’ (ADAMS Accession No.
ML071360456). The changes are consistent
with NRC-approved TS Task Force (TSTF)
Standard TS change TSTF–425, ‘‘Relocate
Surveillance Frequencies to Licensee
Control-Risk Informed Technical
Specifications Task Force (RITSTF) Initiative
5b,’’ Revision 3 (ADAMS Accession No.
ML090850642). The Federal Register notice
published on July 6, 2009 (74 FR 31996),
announced the availability of this TSTF
improvement, and included a model no
significant hazards consideration and safety
evaluation.
Basis for proposed no significant hazards
consideration determination: An analysis of
the no significant hazards consideration was
presented in the TSTF–425. The licensee has
affirmed its applicability of the model no
significant hazards consideration, which is
presented below:
1. Does the proposed change involve
a significant increase in the probability
or consequences of any accident
previously evaluated?
Response: No.
The proposed change relocates the
specified frequencies for periodic
surveillance requirements to licensee control
under a new Surveillance Frequency Control
Program. Surveillance frequencies are not an
initiator to any accident previously
evaluated. As a result, the probability of any
accident previously evaluated is not
significantly increased. The systems and
components required by the technical
specifications for which the surveillance
frequencies are relocated are still required to
be operable, meet the acceptance criteria for
the surveillance requirements, and be
capable of performing any mitigation
function assumed in the accident analysis.
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 changes
do not involve a physical alteration of the
plant (i.e., no new or different type of
equipment will be installed) or a change in
the methods governing normal plant
operation. In addition, the changes do not
impose any new or different requirements.
The changes do not alter assumptions made
in the safety analysis. The proposed changes
are consistent with the safety analysis
assumptions and current plant operating
practice.
Therefore, the proposed changes do not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
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3. Does the proposed change involve a
significant reduction in the margin of safety?
Response: No.
The design, operation, testing methods,
and acceptance criteria for systems,
structures, and components (SSCs), specified
in applicable codes and standards (or
alternatives approved for use by the NRC)
will continue to be met as described in the
plant licensing basis (including the final
safety analysis report and bases to TS), since
these are not affected by changes to the
surveillance frequencies. Similarly, there is
no impact to safety analysis acceptance
criteria as described in the plant licensing
basis. To evaluate a change in the relocated
surveillance frequency, PPL will perform a
risk evaluation using the guidance contained
in NRC approved NEI 04–10, Rev. 1 in
accordance with the TS SFCP. NEI 04–10,
Rev. 1, methodology provides reasonable
acceptance guidelines and methods for
evaluating the risk increase of proposed
changes to surveillance frequencies
consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not
involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Bryan A. Snapp,
Esquire, Assoc. General Counsel, PPL
Services Corporation, 2 North Ninth St.,
GENTW3, Allentown, PA 18101–1179.
NRC Branch Chief: Douglas A.
Broaddus.
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Southern Nuclear Operating Company,
Inc. (SNC), Docket Nos. 50–424 and 50–
425, Vogtle Electric Generating Plant,
Units 1 and 2, Burke County, Georgia
Date of amendment request: July 18,
2014. A publicly-available version is in
ADAMS under Accession Package No.
ML14203A124.
Description of amendment request:
The licensee requested 23 revisions to
the Technical Specifications (TSs).
These revisions adopt various
previously NRC-approved Technical
Specifications Task Force (TSTF)
Travelers. A list of the requested
revisions is included in Enclosure 1 of
the application.
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 for each of the 24 changes
requested, which is presented below:
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1: TSTF–2–A, Revision 1, ‘‘Relocate the 10
Year Sediment Cleaning of the Fuel Oil
Storage Tank to Licensee Control’’ for TS
pages 3.8.3–3 and 3.8.3–4
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change removes the
Surveillance Requirement for performing
sediment cleaning of diesel fuel oil storage
tanks every 10 years from the Technical
Specifications and places it under licensee
control. Diesel fuel oil storage tank cleaning
is not an initiator of any accident previously
evaluated. This change will have no effect on
diesel generator fuel oil quality, which is
tested in accordance with other Technical
Specifications requirements. Removing the
diesel fuel oil storage tank sediment cleaning
requirements from the Technical
Specifications will have no effect on the
ability to mitigate an accident.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change removes the
requirement to clean sediment from the
diesel fuel oil storage tank from the
Technical Specifications and places it under
licensee control. The margin of safety
provided by the fuel oil storage tank
sediment cleaning is unaffected by this
relocation because the quality of diesel fuel
oil is tested in accordance with other
Technical Specifications requirements.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
2: TSTF–27–A, Revision 3, ‘‘Revise SR
[Surveillance Requirement] Frequency for
Minimum Temperature for Criticality’’ for TS
3.4.2, TS Page 3.4.2–1
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
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The proposed change revises the
Surveillance Frequency for monitoring
[reactor coolant system] RCS temperature to
ensure the minimum temperature for
criticality is met. The Frequency is changed
from a 30 minute Frequency when certain
conditions are met to a periodic Frequency
that it is controlled in accordance with the
Surveillance Frequency Control Program.
The initial Frequency for this Surveillance
will be 12 hours. This will ensure that Tavg
[average temperature] is logged at appropriate
intervals (in addition to strip chart recorders
and computer logging of temperature). The
measurement of RCS temperature is not an
initiator of any accident previously
evaluated. The minimum RCS temperature
for criticality is not changed. As a result, the
mitigation of any accident previously
evaluated is not affected.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change revises the
Surveillance Frequency for monitoring RCS
temperature to ensure the minimum
temperature for criticality is met. The
current, condition based Frequency
represents a distraction to the control room
operator during the critical period of plant
startup. RCS temperature is closely
monitored by the operator during the
approach to criticality, and temperature is
recorded on charts and computer logs.
Allowing the operator to monitor
temperature as needed by the situation and
logging RCS temperature at a periodic
Frequency that it is controlled in accordance
with the Surveillance Frequency Control
Program is sufficient to ensure that the LCO
[Limiting Condition for Operation] is met
while eliminating a diversion of the
operator’s attention.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
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3: TSTF–28–A, Revision 0, ‘‘Delete
Unnecessary Action to Measure Gross
Specific Activity, TS 3.4.16,’’ TS page 3.4–16
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 Required
Action B.1 of Specification 3.4.16, ‘‘RCS
Specific Activity,’’ which requires verifying
that Dose Equivalent I–131 specific activity is
within limits. Determination of Dose
Equivalent I–131 is not an initiator of any
accident previously evaluated. Determination
of Dose Equivalent I–131 has no effect on the
mitigation of any accident previously
evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change eliminates a
Required Action. The activities performed
under the Required Action will still be
performed to determine if the LCO is met or
the plant will exit the Applicability of the
Specification. In either case, the presence of
the Required Action does not provide any
significant margin of safety.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
4: TSTF–45–A, Revision 2, ‘‘Exempt
Verification of CIVs that are Locked, Sealed
or Otherwise Secured,’’ TS 3.6.3, TS pages
3.6.3–4, 3.6.3–5
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change exempts containment
isolation valves (CIVs) located inside and
outside of containment that are locked,
sealed, or otherwise secured in position from
the periodic verification of valve position
required by Surveillance Requirements
3.6.3.3 and 3.6.2.4. The exempted valves are
verified to be in the correct position upon
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being locked, sealed, or secured. Because the
valves are in the condition assumed in the
accident analysis, the proposed change will
not affect the initiators or mitigation of any
accident previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change replaces the periodic
verification of valve position with
verification of valve position followed by
locking, sealing, or otherwise securing the
valve in position. Periodic verification is also
effective in detecting valve mispositioning.
However, verification followed by securing
the valve in position is effective in
preventing valve mispositioning. Therefore,
the proposed change does not involve a
significant reduction in a margin of safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
5: TSTF–46–A, Revision 1, ‘‘Clarify the CIV
Surveillance to Apply Only to Automatic
Isolation Valves,’’ TS 3.6.3, TS page 3.6.3.5
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change revises the
requirements in Technical Specification SR
3.6.3.5, and the associated Bases, to delete
the requirement to verify the isolation time
of ‘‘each power operated’’ containment
isolation valve (CIV) and only require
verification of closure time for each
‘‘automatic power operated isolation valve.’’
The closure times for CIVs that do not receive
an automatic closure signal are not an
initiator of any design basis accident or
event, and therefore the proposed change
does not increase the probability of any
accident previously evaluated. The CIVs are
used to respond to accidents previously
evaluated. Power operated CIVs that do not
receive an automatic closure signal are not
assumed to close in a specified time. The
proposed change does not change how the
plant would mitigate an accident previously
evaluated.
Therefore, the proposed change does not
involve a significant increase in the
PO 00000
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Fmt 4703
Sfmt 4703
11481
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 does not result in a
change in the manner in which the CIVs
provide plant protection or introduce any
new or different operational conditions.
Periodic verification that the closure times
for CIVs that receive an automatic closure
signal are within the limits established by the
accident analysis will continue to be
performed under SR 3.6.3.5. The change does
not alter assumptions made in the safety
analysis, and is consistent with the safety
analysis assumptions and current plant
operating practice. There are also no design
changes associated with the proposed
changes, and the change does not involve a
physical alteration of the plant (i.e., no new
or different type of equipment will be
installed).
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change provides clarification
that only CIVs that receive an automatic
isolation signal are within the scope of the
SR 3.6.3.5. The proposed change does not
result in a change in the manner in which the
CIVs provide plant protection. Periodic
verification that closure times for CIVs that
receive an automatic isolation signal are
within the limits established by the accident
analysis will continue to be performed. The
proposed change does not affect the safety
analysis acceptance criteria for any analyzed
event, nor is there a change to any safety
analysis limit. The proposed change does not
alter the manner in which safety limits,
limiting safety system settings or limiting
conditions for operation are determined, nor
is there any adverse effect on those plant
systems necessary to assure the
accomplishment of protection functions. The
proposed change will not result in plant
operation in a configuration outside the
design basis.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
6: TSTF–87–A, Revision 2, ‘‘Revise ‘RTBs
[Reactor Trip Breaker] Open’ and ‘CRDM
[Control Rod Drive Mechanism] Deenergized’ Actions to ‘Incapable of Rod
Withdrawal,’’’ TS 3.4.5, TS Pages 3.4.5–2,
3.4.9–1
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
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This change revises the Required Actions
for LCO 3.4.5, ‘‘RCS Loops—Mode 3,’’
Conditions C.2 and D.1, from ‘‘De-energize
all control rod drive mechanisms,’’ to ‘‘Place
the Rod Control System in a condition
incapable of rod withdrawal.’’ It also revises
LCO 3.4.9, ‘‘Pressurizer,’’ Required Action
A.1, from requiring Reactor Trip Breakers to
be open after reaching MODE 3 to ‘‘Place the
Rod Control System in a condition incapable
of rod withdrawal,’’ and to require full
insertion of all rods. Inadvertent rod
withdrawal can be an initiator for design
basis accidents or events during certain plant
conditions, and therefore must be prevented
under those conditions. The proposed
Required Actions for LCO 3.4.5 and LCO
3.4.9 satisfy the same intent as the current
Required Actions, which is to prevent
inadvertent rod withdrawal when an
applicable Condition is not met, and is
consistent with the assumptions of the
accident analysis. As a result, the proposed
change does not increase the probability of
any accident previously evaluated. The
proposed change does not change how the
plant would mitigate an accident previously
evaluated, as in both the current and
proposed requirements, rod withdrawal is
prohibited.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change provides less
specific, but equivalent, direction on the
manner in which inadvertent control rod
withdrawal is to be prevented when the
Conditions of LCO 3.4.5 and LCO 3.4.9 are
not met. Rod withdrawal will continue to be
prevented when the applicable Conditions of
LCO 3.4.5 and LCO 3.4.9 are met. There are
no design changes associated with the
proposed changes, and the change does not
involve a physical alteration of the plant (i.e.,
no new or different type of equipment will
be installed). The change does not alter
assumptions made in the safety analysis, and
is consistent with the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change provides the
operational flexibility of allowing alternate,
but equivalent, methods of preventing rod
withdrawal when the applicable Conditions
of LCO 3.4.5 and LCO 3.4.9 are met. The
proposed change does not affect the safety
analysis acceptance criteria for any analyzed
event, nor is there a change to any safety
analysis limit. The proposed change does not
alter the manner in which safety limits,
limiting safety system settings or limiting
conditions for operation are determined, nor
is there any adverse effect on those plant
systems necessary to assure the
accomplishment of protection functions. The
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proposed change will not result in plant
operation in a configuration outside the
design basis.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
7: TSTF–95–A, Revision 0, ‘‘Revise
Completion Time for Reducing Power Range
High trip Setpoint from 8 to 72 Hours,’’ TS
3.2.1, TS Pages 3.2.1–1 and 3.2.2–1
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change extends the time
allowed to reduce the Power Range Neutron
Flux—High trip setpoint when Specification
3.2.1, ‘‘Heat Flux Hot Channel Factor,’’ or
Specification 3.2.2, ‘‘Nuclear Enthalpy Rise
Hot Channel Factor,’’ are not within their
limits. Both specifications require a power
reduction followed by a reduction in the
Power Range Neutron Flux—High trip
setpoint. Because reactor power has been
reduced, the reactor core power distribution
limits are within the assumptions of the
accident analysis. Reducing the Power Range
Neutron Flux—High trip setpoints ensures
that reactor power is not inadvertently
increased. Reducing the Power Range
Neutron Flux—High trip setpoints is not an
initiator to any accident previously
evaluated. The consequences of any accident
previously evaluated with the Power Range
Neutron Flux—High trip setpoints not
reduced are no different under the proposed
Completion Time than under the existing
Completion Time. Therefore, the proposed
change does not involve a significant
increase in the probability or consequences
of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change provides additional
time before requiring the Power Range
Neutron Flux—High trip setpoint be reduced
when the reactor core power distribution
limits are not met. The manual reduction in
reactor power required by the specifications
provides the necessary margin of safety for
this condition. Reducing the Power Range
Neutron Flux—High trip setpoints carries an
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Fmt 4703
Sfmt 4703
increased risk of a reactor trip. Delaying the
trip setpoint reduction until the power
reduction has been completed and the
condition is verified will minimize overall
plant risk.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
8: TSTF–110–A, Revision 2, ‘‘Delete SR
Frequencies Based on Inoperable Alarms,’’
TS 3.1, TS pages 3.1.4–3, 3.1.6–3, 3.2.3–1,
3.2.4–4
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change removes surveillance
Frequencies associated with inoperable
alarms (rod position deviation monitor, rod
insertion limit monitor, AFD [Axial Flux
Difference] monitor and QPTR [Quadrant
Power Tilt Ratio] alarm) from the Technical
Specifications and places the actions in plant
administrative procedures. The subject plant
alarms are not an initiator of any accident
previously evaluated. The subject plant
alarms are not used to mitigate any accident
previously evaluated, as the control room
indications of these parameters are sufficient
to alert the operator of an abnormal condition
without the alarms. The alarms are not
credited in the accident analysis.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change removes surveillance
Frequencies associated with inoperable
alarms (rod position deviation monitor, rod
insertion limit monitor, AFD monitor and
QPTR alarm) from the Technical
Specifications and places the actions in plant
administrative procedures. The alarms are
not being removed from the plant. The
actions to be taken when the alarms are not
available are proposed to be controlled under
licensee administrative procedures. As a
result, plant operation is unaffected by this
change and there is no effect on a margin of
safety.
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Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
mstockstill on DSK4VPTVN1PROD with NOTICES
9: TSTF–142–A, Revision 0, ‘‘Increase the
Completion Time When the Core Reactivity
Balance is Not Within Limit,’’ TS 3.1.2, TS
Page 3.1.2–1
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change extends the
Completion Time to take the Required
Actions when measured core reactivity is not
within the specified limit of the predicted
values. The Completion Time to respond to
a difference between predicted and measured
core reactivity is not an initiator to any
accident previously evaluated. The
consequences of an accident during the
proposed Completion Time are no different
from the consequences of an accident during
the existing Completion Time. Therefore, the
proposed change does not involve a
significant increase in the probability or
consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis. Therefore, the
proposed change does not create the
possibility of a new or different kind of
accident from any accident previously
evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change provides additional
time to investigate and to implement
appropriate operating restrictions when
measured core reactivity is not within the
specified limit of the predicted values. The
additional time will not have a significant
effect on plant safety due to the
conservatisms used in designing the reactor
core and performing the safety analyses and
the low probability of an accident or
transient which would approach the core
design limits during the additional time.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
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10: TSTF–234–A, Revision 1, ‘‘Add Action
for More Than One [D]RPI Inoperable,’’ TS
3.1.7, TS Pages 3.1.7–1 and 3.1.7–2.
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change provides a Condition
and Required Actions for more than one
inoperable digital rod position indicator
(DRPI) per rod group. The DRPIs are not an
initiator of any accident previously
evaluated. The DRPIs are one indication used
by operators to verify control rod insertion
following an accident, however other
indications are available. Therefore, allowing
a finite period to time to correct more than
one inoperable DRPI prior to requiring a
plant shutdown will not result in a
significant increase in the consequences of
any accident previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change provides time to
correct the condition of more than one DRPI
inoperable in a rod group. Compensatory
measures are required to verify that the rods
monitored by the inoperable DRPIs are not
moved to ensure that there is no effect on
core reactivity. Requiring a plant shutdown
with inoperable rod position indications
introduces plant risk and should not be
initiated unless the rod position indication
cannot be repaired in a reasonable period of
time. As a result, the safety benefit provided
by the proposed Condition offsets the small
decrease in safety resulting from continued
operation with more than one inoperable
DRPI.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
11: TSTF–245–A, Revision 1, ‘‘AFW Train
Operable When in Service,’’ TS 3.7.5, TS
Page 3.7.5–3
1. Does the proposed amendment involve
a significant increase in the probability or
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11483
consequences of an accident previously
evaluated?
Response: No.
The proposed change revises the
requirements in Technical Specification
3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’
to clarify the operability of an AFW train
when it is aligned for manual steam generator
level control. The AFW System is not an
initiator of any design basis accident or
event, and therefore the proposed change
does not increase the probability of any
accident previously evaluated. The AFW
System is used to respond to accidents
previously evaluated. The proposed change
does not affect the design of the AFW
System, and no physical changes are made to
the plant. The proposed change does not
significantly change how the plant would
mitigate an accident previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a
change in the manner in which the AFW
System provides plant protection. The AFW
System will continue to supply water to the
steam generators to remove decay heat and
other residual heat by delivering at least the
minimum required flow rate to the steam
generators. There are no design changes
associated with the proposed changes, and
the change does not involve a physical
alteration of the plant (i.e., no new or
different type of equipment will be installed).
The change does not alter assumptions made
in the safety analysis, and is consistent with
the safety analysis assumptions and current
plant operating practice. Manual control of
AFW level control valves is not an accident
initiator.
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?
Responses: No.
The proposed change provides the
operational flexibility of allowing an AFW
train(s) to be considered operable when it is
not in the normal standby alignment and is
temporarily incapable of automatic initiation,
such as during alignment and operation for
manual steam generator level control,
provided it is capable of being manually
realigned to the AFW heat removal mode of
operation. The proposed change does not
result in a change in the manner in which the
AFW System provides plant protection. The
AFW System will continue to supply water
to the steam generators to remove decay heat
and other residual heat by delivering at least
the minimum required flow rate to the steam
generators. The proposed change does not
affect the safety analysis acceptance criteria
for any analyzed event, nor is there a change
to any safety analysis limit. The proposed
change does not alter the manner in which
safety limits, limiting safety system settings
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or limiting conditions for operation are
determined, nor is there any adverse effect on
those plant systems necessary to assure the
accomplishment of protection functions. The
proposed change will not result in plant
operation in a configuration outside the
design basis.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
12: TSTF–247–A, Revision 0, ‘‘Provide
Separate Condition Entry for Each [Power
Operated Relief Valve] PORV and Block
Valve,’’ TS 3.4.11, TS Pages 3.4.11–1, 3.4.11–
2, 3.4.11–3
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change revises the
requirements in Technical Specification
3.4.11, ‘‘Pressurizer PORVs,’’ to clarify that
separate Condition entry is allowed for each
block valve. Additionally, the Actions are
modified to no longer require that the PORVs
be placed in manual operation when both
block valves are inoperable and cannot be
restored to operable status within the
specified Completion Time. This preserves
the overpressure protection capabilities of
the PORVs. The pressurizer block valves are
used to isolate their respective PORV in the
event it is experiencing excessive leakage,
and are not an initiator of any design basis
accident or event. Therefore the proposed
change does not increase the probability of
any accident previously evaluated. The
PORV and block valves are used to respond
to accidents previously evaluated. The
proposed change does not affect the design
of the PORV and block valves, and no
physical changes are made to the plant. The
proposed change does not change how the
plant would mitigate an accident previously
evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a
change in the manner in which the PORV
and block valves provide plant protection.
The PORVs will continue to provide
overpressure protection, and the block valves
will continue to provide isolation capability
in the event a PORV is experiencing
excessive leakage. There are no design
changes associated with the proposed
changes, and the change does not involve a
physical alteration of the plant (i.e., no new
or different type of equipment will be
installed). The change does not alter
assumptions made in the safety analysis, and
is consistent with the safety analysis
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assumptions and current plant operating
practice. Operation of the PORV block valves
is not an accident initiator.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed changes provide clarification
that separate Condition entry is allowed for
each block valve. Additionally, the Actions
are modified to no longer require that the
PORVs be placed in manual operation when
both block valves are inoperable and cannot
be restored to operable status within the
specified Completion Time. This preserves
the overpressure protection capabilities of
the PORVs. The proposed change does not
result in a change in the manner in which the
PORV and block valves provide plant
protection. The PORVs will continue to
provide overpressure protection, and the
block valves will continue to provide
isolation capability in the event a PORV is
experiencing excessive leakage. The
proposed change does not affect the safety
analysis acceptance criteria for any analyzed
event, nor is there a change to any safety
analysis limit. The proposed change does not
alter the manner in which safety limits,
limiting safety system settings or limiting
conditions for operation are determined, nor
is there any adverse effect on those plant
systems necessary to assure the
accomplishment of protection functions. The
proposed change will not result in plant
operation in a configuration outside the
design basis.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
13: TSTF–248–A, Revision 0, ‘‘Revise
Shutdown Margin Definition for Stuck Rod
Exception,’’ TS 1.1, TS Page 1.1–6
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change modifies the
definition of Shutdown Margin to eliminate
the requirement to assume the highest worth
control rod is fully withdrawn when
calculating Shutdown Margin if it can be
verified by two independent means that all
control rods are inserted. The method for
calculating shutdown margin is not an
initiator of any accident previously
evaluated. If it can be verified by two
independent means that all control rods are
inserted, the calculated Shutdown Margin
without the conservatism of assuming the
highest worth control rod is withdrawn is
accurate and consistent with the assumptions
in the accident analysis. As a result, the
mitigation of any accident previously
evaluated is not affected.
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Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change modifies the
definition of Shutdown Margin to eliminate
the requirement to assume the highest worth
control rod is fully withdrawn when
calculating Shutdown Margin if it can be
verified by two independent means that all
control rods are inserted. The additional
margin of safety provided by the assumption
that the highest worth control rod is fully
withdrawn is unnecessary if it can be
independently verified that all controls rods
are inserted.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
14: TSTF–266–A, Revision 3, ‘‘Eliminate the
Remote Shutdown System Table of
Instrumentation and Controls,’’ TS 3.3.4, TS
Pages 3.3.4–1, 3.3.4–3
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change removes the list of
Remote Shutdown System instrumentation
and controls from the Technical
Specifications and places them in the Bases.
The Technical Specifications continue to
require that the instrumentation and controls
be operable. The location of the list of
Remote Shutdown System instrumentation
and controls is not an initiator to any
accident previously evaluated. The proposed
change will have no effect on the mitigation
of any accident previously evaluated because
the instrumentation and controls continue to
be required to be operable.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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.
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The proposed change does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change removes the list of
Remote Shutdown System instrumentation
and controls from the Technical
Specifications and places it in the Bases. The
review performed by the NRC when the list
of Remote Shutdown System instrumentation
and controls is revised will no longer be
needed unless the criteria in 10 CFR 50.59
are not met such that prior NRC review is
required. The Technical Specification
requirement that the Remote Shutdown
System be operable, the definition of
operability, the requirements of 10 CFR
50.59, and the Technical Specifications Bases
Control Program are sufficient to ensure that
revision of the list without prior NRC review
and approval does not introduce a significant
safety risk.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
15: TSTF–272–A, Revision 1, ‘‘Refueling
Boron Concentration Clarification,’’ TS 3.9.1,
TS Page 3.9.1–1
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change modifies the
Applicability of Specification 3.9.1, ‘‘Boron
Concentration,’’ to clarify that the boron
concentration limits are only applicable to
the refueling canal and the refueling cavity
when those volumes are attached to the
Reactor Coolant System (RCS). The boron
concentration of water volumes not
connected to the RCS are not an initiator of
an accident previously evaluated. The ability
to mitigate any accident previously evaluated
is not affected by the boron concentration of
water volumes not connected to the RCS.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
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changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change modifies the
Applicability of Specification 3.9.1, ‘‘Boron
Concentration,’’ to clarify that the boron
concentration limits are only applicable to
the refueling canal and the refueling cavity
when those volumes are attached to the RCS.
Technical Specification SR 3.0.4 requires that
Surveillances be met prior to entering the
Applicability of a Specification. As a result,
the boron concentration of the refueling
cavity or the refueling canal must be verified
to satisfy the LCO prior to connecting those
volumes to the RCS. The margin of safety
provided by the refueling boron
concentration is not affected by this change
as the RCS boron concentration will continue
to satisfy the LCO.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
16: TSTF–273–A, Revision 2, ‘‘Safety
Function Determination Program
Clarifications,’’ TS 5.5.15, TS Page 5.5–15
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed TS changes add explanatory
text to the programmatic description of the
Safety Function Determination Program
(SFDP) in Specification 5.5.15 to clarify in
the requirements that consideration does not
have to be made for a loss of power in
determining loss of function. The Bases for
LCO 3.0.6 is revised to provide clarification
of the ‘‘appropriate LCO for loss of function,’’
and that consideration does not have to be
made for a loss of power in determining loss
of function. The changes are editorial and
administrative in nature, and therefore do not
increase the probability of any accident
previously evaluated. No physical or
operational changes are made to the plant.
The proposed change does not change how
the plant would mitigate an accident
previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed changes are editorial and
administrative in nature and do not result in
a change in the manner in which the plant
operates. The loss of function of any specific
component will continue to be addressed in
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11485
its specific TS LCO and plant configuration
will be governed by the required actions of
those LCOs. The proposed changes are
clarifications that do not degrade the
availability or capability of safety related
equipment, and therefore do not create the
possibility of a new or different kind of
accident from any accident previously
evaluated. There are no design changes
associated with the proposed changes, and
the changes do not involve a physical
alteration of the plant (i.e., no new or
different type of equipment will be installed).
The changes do not alter assumptions made
in the safety analysis, and are consistent with
the safety analysis assumptions and current
plant operating practice. Due to the
administrative nature of the changes, they
cannot be an accident initiator.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed changes to TS 5.5.15 are
clarifications and are editorial and
administrative in nature. No changes are
made the LCOs for plant equipment, the time
required for the TS Required Actions to be
completed, or the out of service time for the
components involved. The proposed changes
do not affect the safety analysis acceptance
criteria for any analyzed event, nor is there
a change to any safety analysis limit. The
proposed changes do not alter the manner in
which safety limits, limiting safety system
settings or limiting conditions for operation
are determined, nor is there any adverse
effect on those plant systems necessary to
assure the accomplishment of protection
functions. The proposed changes will not
result in plant operation in a configuration
outside the design basis.
Therefore, the proposed changes do not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
17: TSTF–284–A, Revision 3, ‘‘Add ‘Met vs.
Perform’ to Technical Specification 1.4,
Frequency,’’ TS 1.4, TS 3.4, TS 3.9, TS Pages
1.4–1, 1.4–4, 3.4.11–3, 3.4.12–4 and 3.9.4–2
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed changes insert a discussion
paragraph into Specification 1.4, and several
new examples are added to facilitate the use
and application of SR Notes that utilize the
terms ‘‘met’’ and ‘‘perform.’’ The changes
also modify SRs in multiple Specifications to
appropriately use ‘‘met’’ and ‘‘perform’’
exceptions. The changes are administrative
in nature because they provide clarification
and correction of existing expectations, and
therefore the proposed change does not
increase the probability of any accident
previously evaluated. No physical or
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operational changes are made to the plant.
The proposed change does not significantly
change how the plant would mitigate an
accident previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed changes are administrative
in nature and do not result in a change in the
manner in which the plant operates. The
proposed changes provide clarification and
correction of existing expectations that do
not degrade the availability or capability of
safety related equipment, and therefore do
not create the possibility of a new or different
kind of accident from any accident
previously evaluated. There are no design
changes associated with the proposed
changes, and the changes do not involve a
physical alteration of the plant (i.e., no new
or different type of equipment will be
installed). The changes do not alter
assumptions made in the safety analysis, and
are consistent with the safety analysis
assumptions and current plant operating
practice. Due to the administrative nature of
the changes, they cannot be an accident
initiator.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed changes are administrative
in nature and do not result in a change in the
manner in which the plant operates. The
proposed changes provide clarification and
correction of existing expectations that do
not degrade the availability or capability of
safety related equipment, or alter their
operation. The proposed changes do not
affect the safety analysis acceptance criteria
for any analyzed event, nor is there a change
to any safety analysis limit. The proposed
changes do not alter the manner in which
safety limits, limiting safety system settings
or limiting conditions for operation are
determined, nor is there any adverse effect on
those plant systems necessary to assure the
accomplishment of protection functions. The
proposed changes will not result in plant
operation in a configuration outside the
design basis.
Therefore, the proposed changes do not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
18: TSTF–308–A, Revision 1, ‘‘Determination
of Cumulative and Projected Dose
Contributions in RECP [Radioactive Effluent
Controls Program],’’ TS 5.5.4, TS Page 5.5–3
1. Does the proposed amendment involve
a significant increase in the probability or
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consequences of an accident previously
evaluated?
Response: No.
The proposed change revises Specification
5.5.4, ‘‘Radioactive Effluent Controls
Program,’’ paragraph e, to describe the
original intent of the dose projections. The
cumulative and projection of doses due to
liquid releases are not an assumption in any
accident previously evaluated and have no
effect on the mitigation of any accident
previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change revises Specification
5.5.4, ‘‘Radioactive Effluent Controls
Program,’’ paragraph e, to describe the
original intent of the dose projections. The
cumulative and projection of doses due to
liquid releases are administrative tools to
assure compliance with regulatory limits.
The proposed change revises the requirement
to clarify the intent, thereby improving the
administrative control over this process. As
a result, any effect on the margin of safety
should be minimal.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
19: TSTF–312–A, Revision 1,
‘‘Administrative Control of Containment
Penetrations,’’ TS 3.9.4, TS Page 3.9.4–1
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change would allow
containment penetrations to be unisolated
under administrative controls during core
alterations or movement of irradiated fuel
assemblies within containment. The status of
containment penetration flow paths (i.e.,
open or closed) is not an initiator for any
design basis accident or event, and therefore
the proposed change does not increase the
probability of any accident previously
evaluated. The proposed change does not
affect the design of the primary containment,
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or alter plant operating practices such that
the probability of an accident previously
evaluated would be significantly increased.
The proposed change does not significantly
change how the plant would mitigate an
accident previously evaluated, and is
bounded by the fuel handling accident (FHA)
accident analysis.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
Allowing penetration flow paths to be open
is not an initiator for any accident. The
proposed change to allow open penetration
flow paths will not affect plant safety
functions or plant operating practices such
that a new or different accident could be
created. There are no design changes
associated with the proposed changes, and
the change does not involve a physical
alteration of the plant (i.e., no new or
different type of equipment will be installed).
The change does not alter assumptions made
in the safety analysis, and is consistent with
the safety analysis assumptions and current
plant operating practice.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
TS 3.9.4 provides measures to ensure that
the dose consequences of a postulated FHA
inside containment are minimized. The
proposed change to LCO 3.9.4 will allow
penetration flow path(s) to be open during
refueling operations under administrative
control. These administrative controls will
can and will be achieved in the event of an
FHA inside containment, and will minimize
dose consequences. The proposed change is
bounded by the existing FHA analysis. The
proposed change does not affect the safety
analysis acceptance criteria for any analyzed
event, nor is there a change to any safety
analysis limit. The proposed change does not
alter the manner in which safety limits,
limiting safety system settings or limiting
conditions for operation are determined, nor
is there any adverse effect on those plant
systems necessary to assure the
accomplishment of protection functions. The
proposed change will not result in plant
operation in a configuration outside the
design basis.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
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20: TSTF–314–A, Revision 0, ‘‘Require Static
and Transient FQ Measurement,’’ TS 3.1.4,
3.2.4, TS Pages 3.1.4–2, 3.2.4–1, 3.2.4–3
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change revises the Required
Actions of Specification 3.1.4, ‘‘Rod Group
Alignment Limits,’’ and Specification 3.2.4,
‘‘Quadrant Power Tilt Ratio,’’ to require
measurement of both the steady state and
transient portions of the Heat Flux Hot
Channel Factor, FQ(Z). This change will
ensure that the hot channel factors are within
their limits when the rod alignment limits or
quadrant power tilt ratio are not within their
limits. The verification of hot channel factors
is not an initiator of any accident previously
evaluated. The verification that both the
steady state and transient portion of FQ(Z)
are within their limits will ensure this initial
assumption of the accident analysis is met
should a previously evaluated accident
occur.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change revises the Required
Actions in the Specifications for Rod Group
Alignment Limits and Quadrant Power Tilt
Ratio to require measurement of both the
steady state and transient portions of the
Heat Flux Hot Channel Factor, FQ(Z). This
change is a correction that ensures that the
plant conditions are as assumed in the
accident analysis.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
21: TSTF–340–A, Revision 3, ‘‘Allow 7 Day
Completion Time for a Turbine—Driven
AFW Pump Inoperable,’’ TS 3.7.5, TS Page
3.7.5–1
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
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Response: No.
The proposed change revises Specification
3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’
to allow a 7 day Completion Time to restore
an inoperable AFW turbine-driven pump in
Mode 3 immediately following a refueling
outage, if Mode 2 has not been entered. An
inoperable AFW turbine-driven pump is not
an initiator of any accident previously
evaluated. The ability of the plant to mitigate
an accident is no different while in the
extended Completion Time than during the
existing Completion Time.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in of safety?
Response: No.
The proposed change revises Specification
3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’
to allow a 7-day Completion Time to restore
an inoperable turbine-driven AFW pump in
Mode 3 immediately following a refueling
outage if Mode 2 has not been entered. In
Mode 3 immediately following a refueling
outage, core decay heat is low and the need
for AFW is also diminished. The two
operable motor driven AFW pumps are
available and there are alternate means of
decay heat removal if needed. As a result, the
risk presented by the extended Completion
Time is minimal.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
22: TSTF–343–A, Revision 1, ‘‘Containment
Structural Integrity,’’ TS 5.5, TS Page 5.5–16
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change revises the Technical
Specifications (TS) Administrative Controls
programs for consistency with the
requirements of 10 CFR 50, paragraph
55a(g)(4) for components classified as Code
Class CC. The proposed changes affect the
frequency of visual examinations that will be
performed for the steel containment liner
plate for the purpose of the Containment
Leakage Rate Testing Program.
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The frequency of visual examinations of
the containment and the mode of operation
during which those examinations are
performed does not affect the initiation of
any accident previously evaluated. The use
of NRC approved methods and frequencies
for performing the inspections will ensure
the containment continues to perform the
mitigating function assumed for accidents
previously evaluated.
Therefore, the proposed changes do not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change revises the TS
Administrative Controls programs for
consistency with the requirements of 10 CFR
50, paragraph 55a(g)(4) for components
classified as Code Class CC. The proposed
change affects the frequency of visual
examinations that will be performed for the
steel containment liner plate for the purpose
of the Containment Leakage Rate Testing
Program.
The proposed changes do not involve a
modification to the physical configuration of
the plant (i.e., no new equipment will be
installed) or change in the methods
governing normal plant operation. The
proposed changes will not impose any new
or different requirements or introduce a new
accident initiator, accident precursor, or
malfunction mechanism. Additionally, there
is no change in the types or increases in the
amounts of any effluent that may be released
off-site and there is no increase in individual
or cumulative occupational exposure.
Therefore, the proposed changes do not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed changes revise the Technical
Specifications (TS) Administrative Controls
programs for consistency with the
requirements of 10 CFR 50, paragraph
55a(g)(4) for components classified as Code
Class CC. The proposed change affects the
frequency of visual examinations that will be
performed for the steel containment liner
plate for the purpose of the Containment
Leakage Rate Testing Program. The safety
function of the containment as a fission
product barrier will be maintained.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
23: TSTF–349–A, Revision 1, ‘‘Add Note to
LCO 3.9.5 Allowing Shutdown Cooling
Loops Removal From Operation,’’ TS 3.9.6,
TS Page 3.9.6–1
1. Does the proposed amendment involve
a significant increase in the probability or
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consequences of an accident previously
evaluated?
Response: No.
The proposed change adds an LCO Note to
LCO 3.9.6, ‘‘RHR and Coolant Circulation—
Low Water Level,’’ to allow securing the
operating train of Residual Heat Removal
(RHR) for up to 15 minutes to support
switching operating trains. The allowance is
restricted to conditions in which core outlet
temperature is maintained at least 10 degrees
F below the saturation temperature, when
there are no draining operations, and when
operations that could reduce the reactor
coolant system (RCS) boron concentration are
prohibited. Securing an RHR train to
facilitate the changing of the operating train
is not an initiator to any accident previously
evaluated. The restrictions on the use of the
allowance ensure that an RHR train will not
be needed during the 15 minute period to
mitigate any accident previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of any 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 does not involve a
physical alteration to the plant (i.e., no new
or different type of equipment will be
installed) or a change to the methods
governing normal plant operation. The
changes do not alter the assumptions made
in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed change adds an LCO Note to
LCO 3.9.6, ‘‘RHR and Coolant Circulation—
Low Water Level,’’ to allow securing the
operating train of RHR to support switching
operating trains. The allowance is restricted
to conditions in which core outlet
temperature is maintained at least 10 degrees
F below the saturation temperature, when
there are no draining operations, and when
operations that could reduce the reactor
coolant system (RCS) boron concentration are
prohibited. With these restrictions, combined
with the short time frame allowed to swap
operating RHR trains and the ability to start
an operating RHR train if needed, the
occurrence of an event that would require
immediate operation of an RHR train is
extremely remote.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based on the above, SNC concludes that
the proposed amendment does not involve a
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ‘‘no significant
hazards consideration’’ is justified.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
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standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Leigh D. Perry,
SVP & General Counsel of Operations
and Nuclear, Southern Nuclear
Operating Company, 40 Iverness Center
Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J.
Pascarelli.
Virginia Electric and Power Company,
Docket Nos. 50–338 and 50–339, North
Anna Power, Unit Nos. 1 and 2, Louisa
County, Virginia
Date of amendment request: February
4, 2015. A publicly-available version is
in ADAMS under Accession No.
ML15041A667.
Description of amendment request:
The proposed license amendment
requests the changes to the Technical
Specification (TS) TS 3.1.7, Rod
Position Indication, to provide an
additional monitoring option for an
inoperable control rod position
indicator. Specifically, the proposed
changes would allow monitoring of
control rod drive mechanism stationary
gripper coil voltage every eight hours as
an alternative to using the movable in
core detectors every eight hours to
verify control rod position.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change provides an
alternative method for verifying rod position
of one rod. The proposed change meets the
intent of the current specification in that it
ensures verification of position of the rod
once every 8 hours. The proposed change
provides only an alternative method of
monitoring rod position and does not change
the assumptions or results of any previously
evaluated accident.
Therefore, operation of the facility in
accordance with the proposed amendment
would not involve a significant increase in
the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change provides only an
alternative method of determining the
position of one rod. No new accident
initiators are introduced by the proposed
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alternative manner of performing rod
position verification. The proposed change
does not affect the reactor protection system.
Hence, no new failure modes are created that
would cause a new or different kind of
accidents from any accident previously
evaluated.
Therefore, operation of the facility in
accordance with the proposed amendments
would not create the possibility of a new or
different kind of accident from any
previously evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The basis of TS 3.1.7 states that the
operability of the rod position indicators 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 an alternative method for
determining the position of the affected rod.
As a result, the initial conditions of the
accident analysis are preserved and the
consequences of previously analyzed
accidents are unaffected.
Therefore, operation of the facility in
accordance with the proposed amendment
would not involve a significant reduction in
a margin of safety.
Based on the above, Dominion concludes
that the proposed amendment 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 50.92(c) are satisfied.
Therefore, the NRC staff proposes to
determine that the amendment request
involves no significant hazards
consideration.
Attorney for licensee: Lillian M.
Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar
Street, RS–2, Richmond, VA 23219.
NRC Branch Chief: Robert Pascarelli.
III. Notice of Issuance of Amendments
to Facility Operating Licenses and
Combined Licenses
During the period since publication of
the last biweekly notice, the
Commission has issued the following
amendments. The Commission has
determined for each of these
amendments that the application
complies with the standards and
requirements of the Atomic Energy Act
of 1954, as amended (the Act), and the
Commission’s rules and regulations.
The Commission has made appropriate
findings as required by the Act and the
Commission’s rules and regulations in
10 CFR Chapter I, which are set forth in
the license amendment.
A notice of consideration of issuance
of amendment to facility operating
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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 Energy Kewaunee, Inc.
Docket No. 50–305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment:
May 29, 2013, as supplemented by
letters dated September 23, October 15,
October 17, October 31, and November
7, 2013, and January 7, March 13, April
29, and October 6, 2014, and January 15,
2015.
Brief description of amendment: The
amendment revised the Renewed
Facility Operating License and
associated Technical Specifications to
conform to the permanent shutdown
and defueled status of the facility. It also
denied a proposal to delete paragraphs
1.B, 1.I, and 1.J of the Kewaunee
Operating License.
Date of issuance: February 13, 2015.
Effective date: As of the date of
issuance and shall be implemented
within 90 days from the date of
issuance.
Amendment No.: 215. A publiclyavailable version is in ADAMS under
Accession No. ML14237A045;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Renewed Facility Operating License
No. DPR–43: The amendment revised
the renewed facility operating license
and Technical Specifications.
Date of initial notice in Federal
Register: August 20, 2013 (78 FR
51224). The supplemental letters dated
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September 23, October 15, October 17,
October 31, and November 7, 2013, and
January 7, March 13, April 29, and
October 6, 2014, and January 15, 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
Safety Evaluation dated February 13,
2015.
No significant hazards consideration
comments received: No.
Duke Energy Carolinas, LLC, Docket
Nos. 50–413 and 50–414, Catawba
Nuclear Station, Units 1 and 2, York
County, South Carolina
Duke Energy Carolinas, LLC, Docket
Nos. 50–369 and 50–370 McGuire
Nuclear Station, Units 1 and 2,
Mecklenburg County, North Carolina
Duke Energy Carolinas, LLC, Docket
Nos. 50–269, 50–270, and 50–287,
Oconee Nuclear Station, Units 1, 2, and
3, Oconee County, South Carolina
Date of application for amendments:
July 21, 2014.
Brief description of amendments: The
amendment revises the licensed
operator training requirements to be
consistent with the National Academy
for Nuclear Training (NANT) program.
Additionally, the amendment makes
administrative changes to Technical
Specification Sections 5.1,
‘‘Responsibility;’’ 5.2, ‘‘Organization;’’
5.3, ‘‘Unit Staff Qualifications;’’ 5.5,
‘‘Programs and Manuals;’’ and for
Catawba and McGuire, Section 5.7,
‘‘High Radiation Area.’’
Date of issuance: February 12, 2015.
Effective date: This license
amendment is effective as of its date of
issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 273, 269, 276, 256,
389, 391, and 390. A publicly-available
version is available in ADAMS under
Accession No. ML15002A324.
Renewed Facility Operating License
Nos. NPF–35, NPF–52, NPF–9, NPF–17,
DPR–38, DPR–47, and DPR–55:
Amendments revised the licenses and
Technical Specifications.
Date of initial notice in Federal
Register: November 12, 2014 (79 FR
67199).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated February 12,
2015.
No significant hazards consideration
comments received: No.
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Entergy Operations, Inc., Docket No. 50–
368, Arkansas Nuclear One, Unit No. 2,
Pope County, Arkansas
Date of application for amendment:
December 17, 2012, as supplemented by
letters dated November 7, and December
4, 2013; January 6, May 22, June 30,
August 7, September 24, and December
9, 2014.
Brief description of amendment: The
amendment authorized the transition of
the Arkansas Nuclear One, Unit No. 2,
fire protection program to a riskinformed, performance-based program
based on National Fire Protection
Association (NFPA) 805, in accordance
with 10 CFR 50.48(c). NFPA 805 allows
the use of performance-based methods
such as fire modeling and risk-informed
methods such as fire probabilistic risk
assessment to demonstrate compliance
with the nuclear safety performance
criteria.
Date of issuance: February 18, 2015.
Effective date: As of its date of
issuance and shall be implemented by 6
months from the date of issuance.
Amendment No.: 300. A publiclyavailable version is in ADAMS under
Accession No. ML14356A227;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Renewed Facility Operating License
No. NPR–6: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal
Register: July 23, 2013 (78 FR 44171).
The supplemental letters dated
November 7 and December 4, 2013; and
January 6, May 22, June 30, August 7,
September 24, and December 9, 2014,
provided additional information that
clarified the application, did not expand
the scope of the application as originally
noticed, and did not change the staff’s
original proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated February 18,
2015.
No significant hazards consideration
comments received: No.
Entergy Nuclear FitzPatrick, LLC and
Entergy Nuclear Operations, Inc.,
Docket No. 50–333, James A. FitzPatrick
Nuclear Power Plant, Oswego County,
New York
Date of amendment request: October
8, 2013, as supplemented by a letter
dated November 18, 2014.
Brief description of amendment: The
amendment modifies the Technical
Specifications (TSs) to reduce the
reactor steam dome pressure associated
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with the Reactor Core Safety Limit from
785 psig to 685 psig in TS 2.1.1.1 and
TS 2.1.1.2. This change addresses the
potential to not meet the pressure/
thermal power/minimal critical power
ratio TS safety limit during a pressure
regulator failure-maximum demand
(open) (PRFO) transient. The PRFO
transient was reported by General
Electric as a notification pursuant to
Title 10 of the Code of Federal
Regulations, Part 21, ‘‘Reporting of
Defects and Noncompliance.’’
Date of issuance: February 9, 2015.
Effective date: As of the date of
issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 309. A publiclyavailable version is in ADAMS under
Accession No. ML15014A277;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Renewed Facility Operating License
No. DPR–59: Amendment revised the
Renewed Facility Operating License and
Technical Specifications.
Date of initial notice in Federal
Register: July 8, 2014 (79 FR 38589).
The supplemental letter dated
November 18, 2014, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated February 9,
2015.
No significant hazards consideration
comments received: No.
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Entergy Nuclear Vermont Yankee, LLC
and Entergy Nuclear Operations, Inc.,
Docket No. 50–271, Vermont Yankee
Nuclear Power Station, Vernon,
Vermont
Date of amendment request:
November 14, 2013, as supplemented by
letters dated June 9, 2014, August 6,
2014, and October 9, 2014.
Description of amendment request:
The amendment eliminates operability
requirements for secondary containment
when handling sufficiently decayed
irradiated fuel or a fuel cask following
a minimum of 13 days after the
permanent cessation of reactor
operation.
Date of Issuance: February 12, 2015.
Effective date: The license
amendment becomes effective 13 days
after the licensee’s submittal of the
certifications, as required by 10 CFR
50.82(a)(1)(i) and (ii).
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Amendment No.: 262. A publiclyavailable version is in ADAMS under
Accession No. ML14304A588;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Facility Operating License No. DPR–
28: The amendment revised the
Renewed Facility Operating License and
Technical Specifications.
Date of initial notice in Federal
Register: September 16, 2014 (79 FR
55511).
The supplemental letters dated June
9, 2014, August 6, 2014, and October 9,
2014, 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 this amendment is contained in a
Safety Evaluation dated February 12,
2015.
No significant hazards consideration
comments received: No.
NextEra Energy Duane Arnold, LLC,
Docket No. 50–331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: June 23,
2014.
Brief description of amendment: The
amendment revised the Technical
Specification (TS) requirements to
address NRC Generic Letter (GL) 2008–
01, ‘‘Managing Gas Accumulation in
Emergency Core Cooling, Decay Heat
Removal, and Containment Spray
Systems,’’ as described in TSTF–523,
Revision 2, ‘‘Generic Letter 2008–01,
Managing Gas Accumulation.’’
Date of issuance: February 10, 2015.
Effective date: As of the date of
issuance and shall be implemented
within 90 days.
Amendment No.: 290. A publiclyavailable version is in ADAMS under
Accession No. ML15014A200;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Renewed Facility Operating License
No. DPR–49: The amendment revised
the Renewed Facility Operating License
and Technical Specifications.
Date of initial notice in Federal
Register: September 30, 2014 (79 FR
58820).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated February 10,
2015.
No significant hazards consideration
comments received: No
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NextEra Energy Seabrook, LLC, Docket
No. 50–443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: June 24,
2014, as supplemented by letter dated
December 11, 2014.
Brief description of amendment: The
amendment revised the Seabrook
Technical Specifications (TSs).
Specifically, the amendment modifies
Seabrook TSs to address U.S. Nuclear
Regulatory Commission Generic Letter
(GL) 2008–01, ‘‘Managing Gas
Accumulation in Emergency Core
Cooling, Decay Heat Removal, and
Containment Spray Systems,’’ as
described in TSTF–523, Revision 2,
‘‘Generic Letter 2008–01, Managing Gas
Accumulation.’’
Date of issuance: February 6, 2015.
Effective date: As of its date of
issuance and shall be implemented
within 60 days.
Amendment No.: 144. A publiclyavailable version is in ADAMS under
Accession No. ML14345A288;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Facility Operating License No. NPF–
86: The amendment revised the License
and TS.
Date of initial notice in Federal
Register: September 2, 2014 (79 FR
52066). The supplemental letter dated
December 11, 2014, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated February 6,
2015.
No significant hazards consideration
comments received: No.
South Carolina Electric and Gas
Company, South Carolina Public
Service Authority, Docket No. 50–395,
Virgil C. Summer Nuclear Station, Unit
1, Fairfield County, South Carolina
Date of amendment request:
November 15, 2011, as supplemented by
letters dated November 22, 2011;
January 26 and October 10, 2012;
February 1, April 1, October 14, and
November 26, 2013; January 9, February
25, May 2, May 11, August 14, October
9, and December 11, 2014.
Brief description of amendment: The
amendment authorizes the transition of
the V.C. Summer fire protection
program to a risk-informed,
performance-based program based on
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mstockstill on DSK4VPTVN1PROD with NOTICES
National Fire Protection Association
(NFPA) 805, ‘‘Performance-Based
Standard for Fire Protection for Light
Water Reactor Electric Generating
Plants, 2001 Edition’’ (NFPA 805), in
accordance with 10 CFR 50.48(c).
Date of issuance: February 11, 2015.
Effective date: This amendment is
effective as of its date of issuance and
shall be implemented per the December
11, 2014, supplement, Attachment S,
Table S–2 ‘‘Implementation Items’’,
requiring full implementation by March
31, 2016.
Amendment No.: 199. A publiclyavailable version is in ADAMS under
Accession No. ML14287A289;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Renewed Facility Operating License
No. NPF–12: Amendment revised the
Facility Operating License.
Date of initial notice in Federal
Register: August 14, 2012 (77 FR
48561). The supplemental letters dated
November 22, 2011; October 10, 2012;
February 1, April 1, October 14, and
November 26, 2013; January 9, February
25, May 2, May 11, August 14, October
9, and December 11, 2014, provided
additional information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated February 11,
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–
366, Edwin I. Hatch Nuclear Plant
(HNP), Unit No. 2, Appling County,
Georgia
Date of amendment request: August 8,
2014, as supplemented by letters dated
September 8 and October 24, 2014.
Brief description of amendments: The
amendment revises the Technical
Specification value of the Safety Limit
Minimum Critical Power Ratio to
support operation in the next fuel cycle.
Date of issuance: February 18, 2015.
Effective date: As of the date of
issuance and shall be implemented
prior to reactor startup following the
HNP, Unit 2, spring 2015 refueling
outage.
Amendment No(s).: 218. A publiclyavailable version is in ADAMS under
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Accession No. ML15020A434;
documents related to this amendment
are listed in the Safety Evaluation
enclosed with the amendment.
Renewed Facility Operating License
Nos. DPR–57 and NPF–5: Amendment
revised the licenses and the Technical
Specifications.
Date of initial notice in Federal
Register: January 6, 2015, (80 FR 536).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated February 18,
2015.
No significant hazards consideration
comments received: No.
64546). The supplements dated May 12
(two letters), May 19, and December 17,
2014, provided additional information
that clarified the application, did not
expand the scope of the application as
originally noticed, and did not change
the staff’s original proposed no
significant hazards consideration
determination as published in the
Federal Register.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated February 13,
2015.
No significant hazards consideration
comments received: No.
STP Nuclear Operating Company,
Docket Nos. 50–498 and 50–499, South
Texas Project (STP), Units 1 and 2,
Matagorda County, Texas
Tennessee Valley Authority, Docket No.
50–259, Browns Ferry Nuclear Plant,
Unit 1, Limestone County, Alabama
Date of amendment request:
December 18, 2013, as supplemented by
letter dated June 13, 2014.
Brief description of amendment: The
amendment revised the Technical
Specification (TS) 3.4.9, ‘‘RCS [Reactor
Coolant System] Pressure and
Temperature (P/T) Limits,’’ Figures
3.4.9–1 through 3.4.9–2. The P/T limits
are based on proprietary topical report
NEDC–33178P–A, Revision 1, ‘‘GE
[General Electric] Hitachi Nuclear
Energy Methodology for Development of
Reactor Pressure Vessel PressureTemperature Curves.’’ NEDO–33178–A,
Revision 1 is the non-proprietary
version of the NRC-approved topical
report.
Date of issuance: February 2, 2015.
Effective date: As of the date of
issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 287. A publicly
available version is in ADAMS under
Accession No. ML14325A501;
documents related to this amendment
are listed in the Safety Evaluation (SE)
enclosed with the amendment.
Renewed Facility Operating License
No. DPR–33: Amendment revised the
TSs and the Operating License.
Date of initial notice in Federal
Register: May 6, 2014 (79 FR 25902).
The supplemental letter dated June 13,
2014, provided additional information
that clarified the application, did not
expand the scope of the application as
originally noticed, and did not change
the staff’s original proposed no
significant hazards consideration
determination as published in the
Federal Register.
The Commission’s related evaluation
of the amendment is contained in the SE
dated February 2, 2015.
No significant hazards consideration
comments received: No.
Date of amendment request: July 23,
2013, as supplemented by letters dated
May 12 (two letters), May 19, and
December 17, 2014.
Brief description of amendments: The
amendments revised the STP, Units 1
and 2, Fire Protection Program (FPP)
related to the alternate shutdown
capability. Specifically, it approves the
following operator actions in the control
room prior to evacuation due to a fire
for meeting the alternate shutdown
capability, in addition to manually
tripping the reactor that is currently
credited in the STP, Units 1 and 2, FPP
licensing basis:
• Initiate main steam line isolation
• Closing the pressurizer poweroperated relief valves block valves
• Securing all reactor coolant pumps
• Closing feedwater isolation valves
• Securing the startup feedwater
pump
• Isolating reactor coolant system
letdown
• Securing the centrifugal charging
pumps
In addition, the licensee credits the
automatic trip of the main turbine upon
the initiation of a manual reactor trip for
meeting the alternate shutdown
capability.
Date of issuance: February 13, 2015.
Effective date: As of the date of
issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: Unit 1—203; Unit
2—191. A publicly-available version is
in ADAMS under Accession No.
ML14339A170; documents related to
these amendments are listed in the
Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. NPF–
76 and NPF–80: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal
Register: October 29, 2013 (78 FR
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Dated at Rockville, Maryland, this 23rd day
of February 2015.
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Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2015–04298 Filed 3–2–15; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[NRC–2015–0030]
Applications and Amendments to
Facility Operating Licenses and
Combined Licenses Involving
Proposed No Significant Hazards
Considerations and Containing
Sensitive Unclassified Non-Safeguards
Information and Order Imposing
Procedures for Access to Sensitive
Unclassified Non-Safeguards
Information
Nuclear Regulatory
Commission.
ACTION: License amendment request;
opportunity to comment, request a
hearing, and petition for leave to
intervene; order.
AGENCY:
The U.S. Nuclear Regulatory
Commission (NRC) received and is
considering approval of four
amendment requests. The amendment
requests are for Braidwood Station,
Units 1 and 2, and Byron Station, Units
1 and 2; Peach Bottom Atomic Power
Station, Unit 2; Diablo Canyon Nuclear
Power Plant, Units 1 and 2; and Vogtle
Electric Generating Plant, Units 1 and 2,
Joseph M. Farley Nuclear Plant, Units 1
and 2, and Edwin I. Hatch Nuclear
Plant, Units 1 and 2. The NRC proposes
to determine that each amendment
request involves no significant hazards
consideration. In addition, each
amendment request contains sensitive
unclassified non-safeguards information
(SUNSI).
DATES: Comments must be filed by April
2, 2015. A request for a hearing must be
filed by May 4, 2015. Any potential
party as defined in § 2.4 of Title 10 of
the Code of Federal Regulations (10
CFR), who believes access to SUNSI is
necessary to respond to this notice must
request document access by March 13,
2015.
ADDRESSES: You may submit comments
by any of the following methods (unless
this document describes a different
method for submitting comments on a
specific subject):
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2015–0030. Address
questions about NRC dockets to Carol
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SUMMARY:
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Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov.
• Mail comments to: Cindy Bladey,
Office of Administration, Mail Stop:
O12–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:
Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001; telephone: 301–415–
5411; email: Shirley.Rohrer@nrc.gov.
SUPPLEMENTARY INFORMATION:
submissions at https://
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.
I. Obtaining Information and
Submitting Comments
II. Background
Pursuant to Section 189a.(2) of the
Atomic Energy Act of 1954, as amended
(the Act), the NRC is publishing this
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 notice includes notices of
amendments containing SUNSI.
A. Obtaining Information
Please refer to Docket ID NRC–2015–
0030 when contacting the NRC about
the availability of information for this
action. You may obtain publiclyavailable information related to this
action by any of the following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2015–0030.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publiclyavailable documents online in the
ADAMS Public Documents collection at
https://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 pdr.resource@nrc.gov. The
ADAMS accession number for each
document referenced (if it is available in
ADAMS) is provided the first time that
it is mentioned in the SUPPLEMENTARY
INFORMATION section.
• NRC’s PDR: You may examine and
purchase copies of public documents at
the NRC’s PDR, Room O1–F21, One
White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC–2015–
0030, 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
PO 00000
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III. 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
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
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Agencies
[Federal Register Volume 80, Number 41 (Tuesday, March 3, 2015)]
[Notices]
[Pages 11472-11492]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2015-04298]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2015-0041]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
-----------------------------------------------------------------------
SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission
(NRC) is publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 5, 2015 to February 18, 2015. The
last biweekly notice was published on February 17, 2015.
DATES: Comments must be filed by April 2, 2015. A request for a hearing
must be filed by May 4, 2015.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2015-0041. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov.
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: Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov.
[[Page 11473]]
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2015-0041 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 https://www.regulations.gov and search for Docket ID NRC-2015-0041.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at https://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 pdr.resource@nrc.gov. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2015-0041, 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 will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) 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 https://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
[[Page 11474]]
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 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.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten 10 days prior to the filing deadline, the participant should
contact the Office of the Secretary by email at hearing.docket@nrc.gov,
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 https://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 https://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at https://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 https://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 https://www.nrc.gov/site-help/e-submittals.html, by email to
MSHD.Resource@nrc.gov, 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
[[Page 11475]]
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
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)-(iii).
For further details with respect to 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
No. 2 (ANO-2), Pope County, Arkansas
Date of amendment request: February 6, 2015. A publicly-available
version is in ADAMS under Accession No. ML15041A068.
Description of amendment request: The amendment would revise a Note
to Technical Specification (TS) Surveillance Requirement (SR) 4.1.3.1.2
to exclude Control Element Assembly (CEA) 18 from being exercised per
the SR for the remainder of Cycle 24 due to a degrading upper gripper
coil.
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.
One function of the CEAs is to provide a means of rapid negative
reactivity addition into the core. This occurs upon receipt of a
signal from the Reactor Protection System. This function will
continue to be accomplished with the approval of the proposed
change. Typically, once per 92 days each CEA is moved at least five
inches to ensure the CEA is free to move. CEA 18 remains trippable
(free to move) as illustrated by the last performance of SR
4.1.3.1.2 in January 2015. However, due to abnormally high coil
voltage and current measured on the CEA 18 Upper Gripper Coil (UGC),
future exercising of the CEA could result in the CEA inadvertently
inserting into the core, if the UGC were to fail during the exercise
test. The mis-operation of a CEA, which includes a CEA drop event,
is an abnormal occurrence and has been previously evaluated as part
of the ANO-2 accident analysis. Inadvertent CEA insertion will
result in a reactivity transient and power reduction, and could lead
to a reactor shutdown if the CEA is deemed to be unrecoverable. The
proposed change would minimize the potential for inadvertent
insertion of CEA 18 into the core by maintaining the CEA in place
using the Lower Gripper Coil (LGC), which is operating normally. The
proposed change will not affect the CEAs ability to insert fully
into the core upon receipt of a reactor trip signal.
No modifications are proposed to the Reactor Protection System
or associated Control Element Drive Mechanism Control System logic
with regard to the ability of CEA 18 to remain available for
immediate insertion. The accident mitigation features of the plant
are not affected by the proposed amendment. Because CEA 18 remains
trippable, no additional reactivity considerations need to be taken
into consideration. Nevertheless, Entergy has evaluated the
reactivity consequences associated with failure of CEA 18 to insert
upon a reactor trip in accordance with TS requirements for Shutdown
Margin (SDM) and has determined that SDM requirements would be met
should such an event occur at any time during the remainder of Cycle
24 operation.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
CEA 18 remains trippable. The proposed change will not introduce
any new design changes or systems that can prevent the CEA from
[performing] its specified safety function. As discussed previously,
CEA mis-operation has been previously evaluated in the ANO-2
accident analysis. Furthermore, SDM has been shown to remain within
limits should an event occur at any time during the remainder of
operating Cycle 24 such that CEA 18 fails to insert into the core
upon receipt of a reactor trip signal.
Therefore, 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.
SR 4.1.3.1.2 is intended to verify CEAs are free to move (i.e.,
not mechanically bound). The physical and electrical design of the
CEAs, and past operating experience, provides high confidence that
CEAs remain trippable whether or not exercised during each SR
interval. Eliminating further exercising of CEA 18 for the remainder
of Cycle 24 operation does not directly relate to the potential for
CEA binding to occur. No mechanical binding has been previously
experienced at ANO-2. CEA 18 is contained within a Shutdown CEA
Group and is not used for reactivity control during power maneuvers
(the CEA must remain fully withdrawn at all times when the reactor
is critical). In addition, Entergy has concluded that required SDM
will be maintained should CEA 18 fail to insert following a reactor
trip at any point during the remainder of Cycle 24 operation.
Therefore, this 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 Acting Branch Chief: Eric R. Oesterle.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 1, 2014, as supplemented by
letter dated February 2, 2015. Publicly-available versions are in ADAMS
under Accession Nos. ML14275A374 and ML15033A482.
Description of amendment request: The amendment would relocate
Technical Specifications 3.9.6, ``Refuel Machine,'' and 3.9.7, ``Crane
Travel,'' to the Technical Requirements Manual.
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
[[Page 11476]]
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.
This proposed change relocates Technical Specifications (TS) 3.9.6
(Refuel Machine) and TS 3.9.7 (Crane Travel) to the Waterford 3
Technical Requirements Manual (TRM). This is consistent with the
requirements of [10 CFR 50.36(c)(2)(ii)] and aligns with NUREG-1432
(Combustion Engineering Standard Technical Specifications).
The applicable TS 3.9.6 and TS 3.9.7 design basis accident is the
Fuel Handling Accident (FHA) described in the Updated Final Safety
Analysis Report (UFSAR) Section 15.7.3.4. The limiting FHA results in
all the fuel pins in the dropped and impacted fuel assemblies failing
(472 pins or 236 per assembly). The analysis assumes that a fuel
assembly is dropped as an initial condition and no equipment or
intervention can prevent the initiating condition. The proposed change
was evaluated against [10 CFR 50.36(c)(2)(ii)] criteria and shows no
impact to the lowest functional capability or performance levels of
equipment required for safe operation of the facility because the TS
3.9.6 and TS 3.9.7 requirements do not prevent the accident conditions
from occurring and do not limit the severity of the accident. Since,
the dropped fuel assembly and the impacted fuel assembly are both
already failed in the design basis accident scenario, this change could
not result in a significant increase in the accident consequences. The
TS 3.9.6 and TS 3.9.7 equipment are not required to respond, mitigate,
or terminate any design basis accident, thus this change will not
adversely impact the likelihood or probability of a design basis
accident.
The TS 3.9.6 and TS 3.9.7 requirements do not prevent the accident
conditions from occurring and do not limit the severity of the
accident.
Therefore the TS 3.9.6 and TS 3.9.7 relocation to the TRM would not
cause a significant increase in the accident probability or accident
consequences.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This proposed change relocates TS 3.9.6 (Refuel Machine) and TS
3.9.7 (Crane Travel) to the Waterford 3 TRM. In general, Technical
Specifications are based upon the accident analyses. The accident
analyses assumptions and initial conditions must be protected by the
Technical Specifications. This is a requirement as outlined in [10 CFR
50.36].
[10 CFR 50.36(b)] states the technical specifications will be
derived from the analyses and evaluation included in the safety
analysis report.
[10 CFR 50.36(c)(2)(i)] states that [``]the limiting conditions for
operation are the lowest functional capability or performance levels of
equipment required for safe operation of the facility[. . . .''] [10
CFR 50.36(c)(2)(ii)] provides the four criteria in which any one met
requires a limiting condition for operation. The proposed change
demonstrated that the [10 CFR 50.36(c)(2)(ii)] criteria were not met
and the relocation to the TRM is allowable. By not meeting the [10 CFR
50.36(c)(2)(ii)] criteria for inclusion into the TS means that TS 3.9.6
and TS 3.9.7 do not impact the accident analyses previously evaluated
and would not create the possibility of a new or different kind of
accident.
Specifically, TS 3.9.6 and TS 3.9.7 equipment are not
instrumentation used to detect, and indicate in the control room, a
significant abnormal degradation of the reactor coolant pressure
boundary (Criterion 1). TS 3.9.6 and TS 3.9.7 do not contain a process
variable, design feature, or operating restriction that is an initial
condition of a Design Basis Accident or Transient analysis that either
assumes the failure of or presents a challenge to the integrity of a
fission product barrier (Criterion 2). TS 3.9.6 and TS 3.9.7 does not
contain a structure, system, or component that is part of the primary
success path and which functions or actuates to mitigate a Design Basis
Accident or Transient that either assumes the failure of or presents a
challenge to the integrity of a fission product barrier (Criterion 3).
Lastly, TS 3.9.6 and TS 3.9.7 do not contain a structure, system, or
component which operating experience or probabilistic safety assessment
has shown to be significant to public health and safety (Criterion 4).
TS 3.9.6 and 3.9.7 are not required to meet the lowest functional
capability or performance levels of equipment required for safe
operation of the facility.
Therefore, the accident analyses are not impacted and the
possibility of a new or different kind of accident from any accident
previously evaluated has not changed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed TS 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane Travel)
relocation to the Waterford 3 TRM is administrative in nature because
all requirements will be relocated. Any changes after being relocated
to the Waterford 3 TRM will require that the [10 CFR 50.59] process be
entered ensuring the public health and safety is maintained. By using
the [10 CFR 50.59] process for future changes, the regulatory
requirements ensure that no significant reduction in the margin of
safety occurs.
In addition, the TS 3.9.6 and TS 3.9.7 requirements do not prevent
the design basis accident conditions from occurring and do not limit
the severity of the accident. Thus, TS 3.9.6 and TS 3.9.7 relocation
will not adversely impact the accident analyses and will not cause a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Meena K. Khanna.
Exelon Generation Company, LLC (EGC), Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: November 17, 2014. A publicly available
version is in ADAMS under Accession No. ML14321A744.
Description of amendment request: The proposed amendment would
revise the NMP2 Technical Specification (TS) Allowable Value for the
Main Steam Line Tunnel Lead Enclosure Temperature-High instrumentation
from an ambient temperature dependent (variable setpoint) to ambient
temperature independent (constant Allowable Value). The changes would
delete Surveillance Requirement (SR) 3.3.6.1.2 and revise the Allowable
Value for Function 1.g on Table 3.3.6.1-1, ``Primary Containment
Isolation 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:
[[Page 11477]]
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 involve a significant increase in the
probability or consequences of an accident previously evaluated because
the performance of any equipment credited in the radiological
consequences of an accident is not affected by the change in the leak
detection capability.
The Main Steam Line Tunnel Lead Enclosure Temperature--High is
provided to detect a steam leak in the lead enclosure and provides
diversity to the high flow instrumentation. This function provides a
mitigating action for a steam leak in the Main Steam Line Tunnel Lead
Enclosure, which could lead to a pipe break. This function does not
affect any accident precursors, and the proposed changes do not affect
the leak detection capability. Additionally, the proposed changes do
not degrade the performance of or increase the challenges to any safety
systems assumed to function in the accident analysis.
Therefore, the proposed changes do 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 do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed changes do not add or remove equipment and do not
physically alter the isolation instrumentation. In addition, the Main
Steam Line Tunnel Lead Enclosure LDS [Leak Detection System] is not
utilized in a different manner. The proposed changes do not introduce
any new accident initiators and new failure modes, nor do they reduce
or adversely affect the capabilities of any plant structure, system, or
component to perform their safety function. The Main Steam Line Tunnel
Lead Enclosure LDS will continue to be operated in the same manner.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not involve a significant reduction in a
margin of safety because the changes eliminate the temperature setpoint
dependency on lead enclosure temperature while maintaining the existing
upper AV [Allowable Value] = 175.6[emsp14][deg]F, that was previously
evaluated and approved. There is no adverse impact on the existing
equipment capability as well as associated structures. The increase in
the steam leak rate and associated crack size continues to be well
below the leak rate associated with critical crack size that leads to
pipe break. The proposed changes continue to provide the same level of
protection against a main steam line break as the existing setpoint
values.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Senior Vice President,
Regulatory Affairs, Nuclear, and General Counsel, Exelon Generation
Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Benjamin G. Beasley.
Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: February 20, 2014, as supplemented by
letters dated December 11, 2014, January 13 and January 28, 2015.
Publicly-available in ADAMS under Accession Nos. ML14070A087,
ML14349A333, ML15029A497 and ML15042A122.
Description of amendment request: The NRC staff has previously made
a proposed determination that the amendment request dated February 20,
2014, involves no significant hazards consideration (see 79 FR 42550,
July 22, 2014). Subsequently, by letter dated January 28, 2015, the
licensee provided additional information that expanded the scope of the
amendment request as originally noticed. Accordingly, this notice
supersedes the previous notice in its entirety.
The amendment would revise the Technical Specifications (TSs) by
relocating specific surveillance frequency requirements to a licensee-
controlled program with implementation of Nuclear Energy Institute
(NEI) 04-10 (Revision 1), ``Risk-Informed Technical Specifications
Initiative 5b, Risk-Informed Method for Control of Surveillance
Frequencies'' (ADAMS Accession No. ML071360456). The licensee stated
that the NEI 04-10 methodology provides reasonable acceptance
guidelines and methods for evaluating the risk increase of proposed
changes to surveillance frequencies, consistent with Regulatory Guide
1.177, ``An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Technical Specifications'' (ADAMS Accession No. ML003740176). The
licensee stated that the changes are consistent with NRC-approved
Technical Specification Task Force (TSTF) Improved Standard Technical
Specifications Change Traveler TSTF-425, ``Relocate Surveillance
Frequencies to Licensee Control--RITSTF [Risk-Informed Technical
Specifications Task Force] Initiative 5b,'' Revision 3 (ADAMS Accession
No. ML090850642). The Federal Register notice published on July 6, 2009
(74 FR 31996), announced the availability of TSTF-425, Revision 3. In
the supplement dated January 28, 2015, the licensee requested (1)
additional surveillance frequencies be relocated to the licensee-
controlled program, (2) editorial changes, (3) administrative
deviations from TSTF-425, and (4) other changes resulting from
differences between the St. Lucie Plant TSs and the TSs on which TSTF-
425 was based.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee 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 relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
significantly increased. The systems and components required by the
technical specifications for which the surveillance frequencies are
relocated are still required to be operable, meet the acceptance
criteria for the surveillance requirements, and be capable of
performing any mitigation function assumed in the accident analysis.
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
[[Page 11478]]
probability or consequences of any 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.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis assumptions and current plant operating
practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, FPL will
perform a probabilistic risk evaluation using the guidance contained
in NRC-approved NEI 04-10, Revision 1 in accordance with the TS
Surveillance Frequency Control Program. NEI 04-10, Revision 1,
methodology provides reasonable acceptance guidelines and methods
for evaluating the risk increase of proposed changes to surveillance
frequencies consistent with Regulatory Guide (RG) 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
NRC Branch Chief: Shana R. Helton.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating, Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: November 13, 2014. A publicly-available
version is in ADAMS under Accession No. ML14337A013.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3/4.5.2, ``ECCS [Emergency Core Cooling
System] Subsystems--Tavg [average temperature] Greater Than
or Equal to 350[emsp14][deg]F [degrees Fahrenheit],'' to correct non-
conservative TS requirements. The licensee also requested editorial
changes to the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is presented as follows:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed TS changes involve TS 3.5.2 Action `a', new TS
3.5.2 Action `h', and the provision in SR [Surveillance Requirement]
4.5.2.a to address non-conservative TS requirements. Editorial changes
are also proposed for consistency and clarity. These changes do not
affect any precursors to any accident previously evaluated and
subsequently, will not impact the probability or consequences of an
accident previously evaluated. Furthermore, these changes do not
adversely affect mitigation equipment or strategies.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed TS changes involve TS 3.5.2 Action `a', new TS
3.5.2 Action `h', and the provision in SR 4.5.2.a to address non-
conservative TS requirements. Editorial changes are also proposed for
consistency and clarity. The proposed changes provide better assurance
that the ECCS systems, subsystems, and components are properly aligned
to support safe reactor operation consistent with the licensing basis
requirements. The proposed changes do not introduce new modes of plant
operation and do not involve physical modifications to the plant (no
new or different type of equipment will be installed). There are no
changes in the method by which any safety related plant structure,
system, or component (SSC) performs its specified safety function. As
such, the plant conditions for which the design basis accident analyses
were performed remain valid.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a result
of the proposed change. There will be no adverse effect or challenges
imposed on any SSC as a result of the proposed change.
Therefore, the proposed changes do 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 the
margin of safety?
No. Margin of safety is related to confidence in the ability of the
fission product barriers to perform their accident mitigation
functions. The proposed TS changes involve TS 3.5.2 Action `a', new TS
3.5.2 Action `h', and the provision in SR 4.5.2.a to address non-
conservative TS requirements. Editorial changes are also proposed for
consistency and clarity. The proposed changes provide better assurance
that the ECCS systems, subsystems, and components are properly aligned
to support safe reactor operation consistent with the licensing basis
requirements. The proposed changes do not physically alter any SSC.
There will be no effect on those SSCs necessary to assure the
accomplishment of specified functions. There will be no impact on the
overpower limit, departure from nucleate boiling ratio (DNBR) limits,
loss of cooling accident peak cladding temperature (LOCA PCT), or any
other margin of safety. The applicable radiological dose consequence
acceptance criteria will continue to be met. Therefore, the proposed
changes do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL
33408-0420.
NRC Branch Chief: Shana R. Helton.
Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: February 6, 2015. A publicly-available
version is in ADAMS under Accession No. ML15041A069.
Description of amendment request: The proposed amendments would
modify the technical specifications requirements for unavailable
barriers by adding limiting condition for operation
[[Page 11479]]
(LCO) 3.0.8. The changes are consistent with the NRC approved Technical
Specification Task Force (TSTF) Standard Technical Specification change
TSTF-427, ``Allowance for Non-Technical Specification Barrier
Degradation on Supported System OPERABILITY,'' Revision 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
affirmed the applicability of the model proposed no significant hazards
consideration published on October 3, 2006 (71 FR 58444), ``Notice of
Availability of the Model Safety Evaluation.'' The findings presented
in that evaluation are presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an unavailable barrier if risk is assessed and managed. The
postulated initiating events which may require a functional barrier are
limited to those with low frequencies of occurrence, and the overall TS
system safety function would still be available for the majority of
anticipated challenges. Therefore, the probability of an accident
previously evaluated is not significantly increased, if at all. The
consequences of an accident while relying on the allowance provided by
proposed LCO 3.0.8 are no different than the consequences of an
accident while relying on the TS required actions in effect without the
allowance provided by proposed LCO 3.0.8. 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, this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to an unavailable barrier, if risk is
assessed and managed, 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 will further minimize
possible concerns. Thus, this change does not create the possibility of
a new or different kind of accident from an accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an unavailable
barrier, if risk is assessed and managed. The postulated initiating
events which may require a functional barrier are limited to those with
low frequencies of occurrence, and the overall TS system safety
function would still be available for the majority of anticipated
challenges. The risk impact of the proposed TS changes was assessed
following the three-tiered approach recommended in RG [Regulatory
Guide] 1.177. A bounding risk assessment was performed to justify the
proposed TS changes. This application of LCO 3.0.8 is predicated upon
the licensee's performance of a risk assessment and the management of
plant risk. The net change to the margin of safety is insignificant as
indicated by the anticipated low levels of associated risk (ICCDP
[incremental conditional core damage probability] and ICLERP
[incremental large early release probability]) as shown in Table 1 of
Section 3.1.1 in the Safety Evaluation. Therefore, this change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, Michigan 49106.
NRC Branch Chief: David L. Pelton.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment requests: October 27, 2014. A publicly-
available version is available in ADAMS under Accession No.
ML14317A052.
Description of amendment requests: The proposed amendments will
modify the Susquehanna technical specifications (TS). Specifically,
the proposed amendments will modify the TS by relocating specific
surveillance frequencies to a licensee-controlled program, the
Surveillance Frequency Control Program (SFCP), with implementation
of Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed Technical
Specifications Initiative 5b, Risk-Informed Method for Control of
Surveillance Frequencies'' (ADAMS Accession No. ML071360456). The
changes are consistent with NRC-approved TS Task Force (TSTF)
Standard TS change TSTF-425, ``Relocate Surveillance Frequencies to
Licensee Control-Risk Informed Technical Specifications Task Force
(RITSTF) Initiative 5b,'' Revision 3 (ADAMS Accession No.
ML090850642). The Federal Register notice published on July 6, 2009
(74 FR 31996), announced the availability of this TSTF improvement,
and included a model no significant hazards consideration and safety
evaluation.
Basis for proposed no significant hazards consideration
determination: An analysis of the no significant hazards
consideration was presented in the TSTF-425. The licensee has
affirmed its applicability of the model no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
significantly increased. The systems and components required by the
technical specifications for which the surveillance frequencies are
relocated are still required to be operable, meet the acceptance
criteria for the surveillance requirements, and be capable of
performing any mitigation function assumed in the accident analysis.
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 changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
[[Page 11480]]
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, PPL will
perform a risk evaluation using the guidance contained in NRC
approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-
10, Rev. 1, methodology provides reasonable acceptance guidelines
and methods for evaluating the risk increase of proposed changes to
surveillance frequencies consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Douglas A. Broaddus.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: July 18, 2014. A publicly-available
version is in ADAMS under Accession Package No. ML14203A124.
Description of amendment request: The licensee requested 23
revisions to the Technical Specifications (TSs). These revisions adopt
various previously NRC-approved Technical Specifications Task Force
(TSTF) Travelers. A list of the requested revisions is included in
Enclosure 1 of the application.
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 for each of the 24 changes requested, which is presented
below:
1: TSTF-2-A, Revision 1, ``Relocate the 10 Year Sediment Cleaning of
the Fuel Oil Storage Tank to Licensee Control'' for TS pages 3.8.3-3
and 3.8.3-4
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change removes the Surveillance Requirement for
performing sediment cleaning of diesel fuel oil storage tanks every
10 years from the Technical Specifications and places it under
licensee control. Diesel fuel oil storage tank cleaning is not an
initiator of any accident previously evaluated. This change will
have no effect on diesel generator fuel oil quality, which is tested
in accordance with other Technical Specifications requirements.
Removing the diesel fuel oil storage tank sediment cleaning
requirements from the Technical Specifications will have no effect
on the ability to mitigate an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change removes the requirement to clean sediment
from the diesel fuel oil storage tank from the Technical
Specifications and places it under licensee control. The margin of
safety provided by the fuel oil storage tank sediment cleaning is
unaffected by this relocation because the quality of diesel fuel oil
is tested in accordance with other Technical Specifications
requirements.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
2: TSTF-27-A, Revision 3, ``Revise SR [Surveillance Requirement]
Frequency for Minimum Temperature for Criticality'' for TS 3.4.2, TS
Page 3.4.2-1
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Surveillance Frequency for
monitoring [reactor coolant system] RCS temperature to ensure the
minimum temperature for criticality is met. The Frequency is changed
from a 30 minute Frequency when certain conditions are met to a
periodic Frequency that it is controlled in accordance with the
Surveillance Frequency Control Program. The initial Frequency for
this Surveillance will be 12 hours. This will ensure that
Tavg [average temperature] is logged at appropriate
intervals (in addition to strip chart recorders and computer logging
of temperature). The measurement of RCS temperature is not an
initiator of any accident previously evaluated. The minimum RCS
temperature for criticality is not changed. As a result, the
mitigation of any accident previously evaluated is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises the Surveillance Frequency for
monitoring RCS temperature to ensure the minimum temperature for
criticality is met. The current, condition based Frequency
represents a distraction to the control room operator during the
critical period of plant startup. RCS temperature is closely
monitored by the operator during the approach to criticality, and
temperature is recorded on charts and computer logs. Allowing the
operator to monitor temperature as needed by the situation and
logging RCS temperature at a periodic Frequency that it is
controlled in accordance with the Surveillance Frequency Control
Program is sufficient to ensure that the LCO [Limiting Condition for
Operation] is met while eliminating a diversion of the operator's
attention.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
[[Page 11481]]
3: TSTF-28-A, Revision 0, ``Delete Unnecessary Action to Measure Gross
Specific Activity, TS 3.4.16,'' TS page 3.4-16
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 Required Action B.1 of
Specification 3.4.16, ``RCS Specific Activity,'' which requires
verifying that Dose Equivalent I-131 specific activity is within
limits. Determination of Dose Equivalent I-131 is not an initiator
of any accident previously evaluated. Determination of Dose
Equivalent I-131 has no effect on the mitigation of any accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change eliminates a Required Action. The activities
performed under the Required Action will still be performed to
determine if the LCO is met or the plant will exit the Applicability
of the Specification. In either case, the presence of the Required
Action does not provide any significant margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
4: TSTF-45-A, Revision 2, ``Exempt Verification of CIVs that are
Locked, Sealed or Otherwise Secured,'' TS 3.6.3, TS pages 3.6.3-4,
3.6.3-5
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change exempts containment isolation valves (CIVs)
located inside and outside of containment that are locked, sealed,
or otherwise secured in position from the periodic verification of
valve position required by Surveillance Requirements 3.6.3.3 and
3.6.2.4. The exempted valves are verified to be in the correct
position upon being locked, sealed, or secured. Because the valves
are in the condition assumed in the accident analysis, the proposed
change will not affect the initiators or mitigation of any accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change replaces the periodic verification of valve
position with verification of valve position followed by locking,
sealing, or otherwise securing the valve in position. Periodic
verification is also effective in detecting valve mispositioning.
However, verification followed by securing the valve in position is
effective in preventing valve mispositioning. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
5: TSTF-46-A, Revision 1, ``Clarify the CIV Surveillance to Apply Only
to Automatic Isolation Valves,'' TS 3.6.3, TS page 3.6.3.5
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the requirements in Technical
Specification SR 3.6.3.5, and the associated Bases, to delete the
requirement to verify the isolation time of ``each power operated''
containment isolation valve (CIV) and only require verification of
closure time for each ``automatic power operated isolation valve.''
The closure times for CIVs that do not receive an automatic closure
signal are not an initiator of any design basis accident or event,
and therefore the proposed change does not increase the probability
of any accident previously evaluated. The CIVs are used to respond
to accidents previously evaluated. Power operated CIVs that do not
receive an automatic closure signal are not assumed to close in a
specified time. The proposed change does not change how the plant
would mitigate an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a change in the manner in
which the CIVs provide plant protection or introduce any new or
different operational conditions. Periodic verification that the
closure times for CIVs that receive an automatic closure signal are
within the limits established by the accident analysis will continue
to be performed under SR 3.6.3.5. The change does not alter
assumptions made in the safety analysis, and is consistent with the
safety analysis assumptions and current plant operating practice.
There are also no design changes associated with the proposed
changes, and the change does not involve a physical alteration of
the plant (i.e., no new or different type of equipment will be
installed).
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides clarification that only CIVs that
receive an automatic isolation signal are within the scope of the SR
3.6.3.5. The proposed change does not result in a change in the
manner in which the CIVs provide plant protection. Periodic
verification that closure times for CIVs that receive an automatic
isolation signal are within the limits established by the accident
analysis will continue to be performed. The proposed change does not
affect the safety analysis acceptance criteria for any analyzed
event, nor is there a change to any safety analysis limit. The
proposed change does not alter the manner in which safety limits,
limiting safety system settings or limiting conditions for operation
are determined, nor is there any adverse effect on those plant
systems necessary to assure the accomplishment of protection
functions. The proposed change will not result in plant operation in
a configuration outside the design basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
6: TSTF-87-A, Revision 2, ``Revise `RTBs [Reactor Trip Breaker] Open'
and `CRDM [Control Rod Drive Mechanism] De-energized' Actions to
`Incapable of Rod Withdrawal,''' TS 3.4.5, TS Pages 3.4.5-2, 3.4.9-1
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
[[Page 11482]]
This change revises the Required Actions for LCO 3.4.5, ``RCS
Loops--Mode 3,'' Conditions C.2 and D.1, from ``De-energize all
control rod drive mechanisms,'' to ``Place the Rod Control System in
a condition incapable of rod withdrawal.'' It also revises LCO
3.4.9, ``Pressurizer,'' Required Action A.1, from requiring Reactor
Trip Breakers to be open after reaching MODE 3 to ``Place the Rod
Control System in a condition incapable of rod withdrawal,'' and to
require full insertion of all rods. Inadvertent rod withdrawal can
be an initiator for design basis accidents or events during certain
plant conditions, and therefore must be prevented under those
conditions. The proposed Required Actions for LCO 3.4.5 and LCO
3.4.9 satisfy the same intent as the current Required Actions, which
is to prevent inadvertent rod withdrawal when an applicable
Condition is not met, and is consistent with the assumptions of the
accident analysis. As a result, the proposed change does not
increase the probability of any accident previously evaluated. The
proposed change does not change how the plant would mitigate an
accident previously evaluated, as in both the current and proposed
requirements, rod withdrawal is prohibited.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change provides less specific, but equivalent,
direction on the manner in which inadvertent control rod withdrawal
is to be prevented when the Conditions of LCO 3.4.5 and LCO 3.4.9
are not met. Rod withdrawal will continue to be prevented when the
applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. There are
no design changes associated with the proposed changes, and the
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed). The change
does not alter assumptions made in the safety analysis, and is
consistent with the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides the operational flexibility of
allowing alternate, but equivalent, methods of preventing rod
withdrawal when the applicable Conditions of LCO 3.4.5 and LCO 3.4.9
are met. The proposed change does not affect the safety analysis
acceptance criteria for any analyzed event, nor is there a change to
any safety analysis limit. The proposed change does not alter the
manner in which safety limits, limiting safety system settings or
limiting conditions for operation are determined, nor is there any
adverse effect on those plant systems necessary to assure the
accomplishment of protection functions. The proposed change will not
result in plant operation in a configuration outside the design
basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
7: TSTF-95-A, Revision 0, ``Revise Completion Time for Reducing Power
Range High trip Setpoint from 8 to 72 Hours,'' TS 3.2.1, TS Pages
3.2.1-1 and 3.2.2-1
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the time allowed to reduce the Power
Range Neutron Flux--High trip setpoint when Specification 3.2.1,
``Heat Flux Hot Channel Factor,'' or Specification 3.2.2, ``Nuclear
Enthalpy Rise Hot Channel Factor,'' are not within their limits.
Both specifications require a power reduction followed by a
reduction in the Power Range Neutron Flux--High trip setpoint.
Because reactor power has been reduced, the reactor core power
distribution limits are within the assumptions of the accident
analysis. Reducing the Power Range Neutron Flux--High trip setpoints
ensures that reactor power is not inadvertently increased. Reducing
the Power Range Neutron Flux--High trip setpoints is not an
initiator to any accident previously evaluated. The consequences of
any accident previously evaluated with the Power Range Neutron
Flux--High trip setpoints not reduced are no different under the
proposed Completion Time than under the existing Completion Time.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides additional time before requiring
the Power Range Neutron Flux--High trip setpoint be reduced when the
reactor core power distribution limits are not met. The manual
reduction in reactor power required by the specifications provides
the necessary margin of safety for this condition. Reducing the
Power Range Neutron Flux--High trip setpoints carries an increased
risk of a reactor trip. Delaying the trip setpoint reduction until
the power reduction has been completed and the condition is verified
will minimize overall plant risk.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
8: TSTF-110-A, Revision 2, ``Delete SR Frequencies Based on Inoperable
Alarms,'' TS 3.1, TS pages 3.1.4-3, 3.1.6-3, 3.2.3-1, 3.2.4-4
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change removes surveillance Frequencies associated
with inoperable alarms (rod position deviation monitor, rod
insertion limit monitor, AFD [Axial Flux Difference] monitor and
QPTR [Quadrant Power Tilt Ratio] alarm) from the Technical
Specifications and places the actions in plant administrative
procedures. The subject plant alarms are not an initiator of any
accident previously evaluated. The subject plant alarms are not used
to mitigate any accident previously evaluated, as the control room
indications of these parameters are sufficient to alert the operator
of an abnormal condition without the alarms. The alarms are not
credited in the accident analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change removes surveillance Frequencies associated
with inoperable alarms (rod position deviation monitor, rod
insertion limit monitor, AFD monitor and QPTR alarm) from the
Technical Specifications and places the actions in plant
administrative procedures. The alarms are not being removed from the
plant. The actions to be taken when the alarms are not available are
proposed to be controlled under licensee administrative procedures.
As a result, plant operation is unaffected by this change and there
is no effect on a margin of safety.
[[Page 11483]]
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
9: TSTF-142-A, Revision 0, ``Increase the Completion Time When the Core
Reactivity Balance is Not Within Limit,'' TS 3.1.2, TS Page 3.1.2-1
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the Completion Time to take the
Required Actions when measured core reactivity is not within the
specified limit of the predicted values. The Completion Time to
respond to a difference between predicted and measured core
reactivity is not an initiator to any accident previously evaluated.
The consequences of an accident during the proposed Completion Time
are no different from the consequences of an accident during the
existing Completion Time. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides additional time to investigate and
to implement appropriate operating restrictions when measured core
reactivity is not within the specified limit of the predicted
values. The additional time will not have a significant effect on
plant safety due to the conservatisms used in designing the reactor
core and performing the safety analyses and the low probability of
an accident or transient which would approach the core design limits
during the additional time. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
10: TSTF-234-A, Revision 1, ``Add Action for More Than One [D]RPI
Inoperable,'' TS 3.1.7, TS Pages 3.1.7-1 and 3.1.7-2.
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change provides a Condition and Required Actions
for more than one inoperable digital rod position indicator (DRPI)
per rod group. The DRPIs are not an initiator of any accident
previously evaluated. The DRPIs are one indication used by operators
to verify control rod insertion following an accident, however other
indications are available. Therefore, allowing a finite period to
time to correct more than one inoperable DRPI prior to requiring a
plant shutdown will not result in a significant increase in the
consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides time to correct the condition of
more than one DRPI inoperable in a rod group. Compensatory measures
are required to verify that the rods monitored by the inoperable
DRPIs are not moved to ensure that there is no effect on core
reactivity. Requiring a plant shutdown with inoperable rod position
indications introduces plant risk and should not be initiated unless
the rod position indication cannot be repaired in a reasonable
period of time. As a result, the safety benefit provided by the
proposed Condition offsets the small decrease in safety resulting
from continued operation with more than one inoperable DRPI.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
11: TSTF-245-A, Revision 1, ``AFW Train Operable When in Service,'' TS
3.7.5, TS Page 3.7.5-3
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the requirements in Technical
Specification 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to
clarify the operability of an AFW train when it is aligned for
manual steam generator level control. The AFW System is not an
initiator of any design basis accident or event, and therefore the
proposed change does not increase the probability of any accident
previously evaluated. The AFW System is used to respond to accidents
previously evaluated. The proposed change does not affect the design
of the AFW System, and no physical changes are made to the plant.
The proposed change does not significantly change how the plant
would mitigate an accident previously evaluated. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a change in the manner in
which the AFW System provides plant protection. The AFW System will
continue to supply water to the steam generators to remove decay
heat and other residual heat by delivering at least the minimum
required flow rate to the steam generators. There are no design
changes associated with the proposed changes, and the change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed). The change does not
alter assumptions made in the safety analysis, and is consistent
with the safety analysis assumptions and current plant operating
practice. Manual control of AFW level control valves is not an
accident initiator.
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?
Responses: No.
The proposed change provides the operational flexibility of
allowing an AFW train(s) to be considered operable when it is not in
the normal standby alignment and is temporarily incapable of
automatic initiation, such as during alignment and operation for
manual steam generator level control, provided it is capable of
being manually realigned to the AFW heat removal mode of operation.
The proposed change does not result in a change in the manner in
which the AFW System provides plant protection. The AFW System will
continue to supply water to the steam generators to remove decay
heat and other residual heat by delivering at least the minimum
required flow rate to the steam generators. The proposed change does
not affect the safety analysis acceptance criteria for any analyzed
event, nor is there a change to any safety analysis limit. The
proposed change does not alter the manner in which safety limits,
limiting safety system settings
[[Page 11484]]
or limiting conditions for operation are determined, nor is there
any adverse effect on those plant systems necessary to assure the
accomplishment of protection functions. The proposed change will not
result in plant operation in a configuration outside the design
basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
12: TSTF-247-A, Revision 0, ``Provide Separate Condition Entry for Each
[Power Operated Relief Valve] PORV and Block Valve,'' TS 3.4.11, TS
Pages 3.4.11-1, 3.4.11-2, 3.4.11-3
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the requirements in Technical
Specification 3.4.11, ``Pressurizer PORVs,'' to clarify that
separate Condition entry is allowed for each block valve.
Additionally, the Actions are modified to no longer require that the
PORVs be placed in manual operation when both block valves are
inoperable and cannot be restored to operable status within the
specified Completion Time. This preserves the overpressure
protection capabilities of the PORVs. The pressurizer block valves
are used to isolate their respective PORV in the event it is
experiencing excessive leakage, and are not an initiator of any
design basis accident or event. Therefore the proposed change does
not increase the probability of any accident previously evaluated.
The PORV and block valves are used to respond to accidents
previously evaluated. The proposed change does not affect the design
of the PORV and block valves, and no physical changes are made to
the plant. The proposed change does not change how the plant would
mitigate an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a change in the manner in
which the PORV and block valves provide plant protection. The PORVs
will continue to provide overpressure protection, and the block
valves will continue to provide isolation capability in the event a
PORV is experiencing excessive leakage. There are no design changes
associated with the proposed changes, and the change does not
involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed). The change does not
alter assumptions made in the safety analysis, and is consistent
with the safety analysis assumptions and current plant operating
practice. Operation of the PORV block valves is not an accident
initiator.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes provide clarification that separate
Condition entry is allowed for each block valve. Additionally, the
Actions are modified to no longer require that the PORVs be placed
in manual operation when both block valves are inoperable and cannot
be restored to operable status within the specified Completion Time.
This preserves the overpressure protection capabilities of the
PORVs. The proposed change does not result in a change in the manner
in which the PORV and block valves provide plant protection. The
PORVs will continue to provide overpressure protection, and the
block valves will continue to provide isolation capability in the
event a PORV is experiencing excessive leakage. The proposed change
does not affect the safety analysis acceptance criteria for any
analyzed event, nor is there a change to any safety analysis limit.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined, nor is there any adverse effect on those
plant systems necessary to assure the accomplishment of protection
functions. The proposed change will not result in plant operation in
a configuration outside the design basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
13: TSTF-248-A, Revision 0, ``Revise Shutdown Margin Definition for
Stuck Rod Exception,'' TS 1.1, TS Page 1.1-6
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the definition of Shutdown Margin
to eliminate the requirement to assume the highest worth control rod
is fully withdrawn when calculating Shutdown Margin if it can be
verified by two independent means that all control rods are
inserted. The method for calculating shutdown margin is not an
initiator of any accident previously evaluated. If it can be
verified by two independent means that all control rods are
inserted, the calculated Shutdown Margin without the conservatism of
assuming the highest worth control rod is withdrawn is accurate and
consistent with the assumptions in the accident analysis. As a
result, the mitigation of any accident previously evaluated is not
affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change modifies the definition of Shutdown Margin
to eliminate the requirement to assume the highest worth control rod
is fully withdrawn when calculating Shutdown Margin if it can be
verified by two independent means that all control rods are
inserted. The additional margin of safety provided by the assumption
that the highest worth control rod is fully withdrawn is unnecessary
if it can be independently verified that all controls rods are
inserted.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
14: TSTF-266-A, Revision 3, ``Eliminate the Remote Shutdown System
Table of Instrumentation and Controls,'' TS 3.3.4, TS Pages 3.3.4-1,
3.3.4-3
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change removes the list of Remote Shutdown System
instrumentation and controls from the Technical Specifications and
places them in the Bases. The Technical Specifications continue to
require that the instrumentation and controls be operable. The
location of the list of Remote Shutdown System instrumentation and
controls is not an initiator to any accident previously evaluated.
The proposed change will have no effect on the mitigation of any
accident previously evaluated because the instrumentation and
controls continue to be required to be operable.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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.
[[Page 11485]]
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change removes the list of Remote Shutdown System
instrumentation and controls from the Technical Specifications and
places it in the Bases. The review performed by the NRC when the
list of Remote Shutdown System instrumentation and controls is
revised will no longer be needed unless the criteria in 10 CFR 50.59
are not met such that prior NRC review is required. The Technical
Specification requirement that the Remote Shutdown System be
operable, the definition of operability, the requirements of 10 CFR
50.59, and the Technical Specifications Bases Control Program are
sufficient to ensure that revision of the list without prior NRC
review and approval does not introduce a significant safety risk.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
15: TSTF-272-A, Revision 1, ``Refueling Boron Concentration
Clarification,'' TS 3.9.1, TS Page 3.9.1-1
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the Applicability of Specification
3.9.1, ``Boron Concentration,'' to clarify that the boron
concentration limits are only applicable to the refueling canal and
the refueling cavity when those volumes are attached to the Reactor
Coolant System (RCS). The boron concentration of water volumes not
connected to the RCS are not an initiator of an accident previously
evaluated. The ability to mitigate any accident previously evaluated
is not affected by the boron concentration of water volumes not
connected to the RCS.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change modifies the Applicability of Specification
3.9.1, ``Boron Concentration,'' to clarify that the boron
concentration limits are only applicable to the refueling canal and
the refueling cavity when those volumes are attached to the RCS.
Technical Specification SR 3.0.4 requires that Surveillances be met
prior to entering the Applicability of a Specification. As a result,
the boron concentration of the refueling cavity or the refueling
canal must be verified to satisfy the LCO prior to connecting those
volumes to the RCS. The margin of safety provided by the refueling
boron concentration is not affected by this change as the RCS boron
concentration will continue to satisfy the LCO.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
16: TSTF-273-A, Revision 2, ``Safety Function Determination Program
Clarifications,'' TS 5.5.15, TS Page 5.5-15
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed TS changes add explanatory text to the programmatic
description of the Safety Function Determination Program (SFDP) in
Specification 5.5.15 to clarify in the requirements that
consideration does not have to be made for a loss of power in
determining loss of function. The Bases for LCO 3.0.6 is revised to
provide clarification of the ``appropriate LCO for loss of
function,'' and that consideration does not have to be made for a
loss of power in determining loss of function. The changes are
editorial and administrative in nature, and therefore do not
increase the probability of any accident previously evaluated. No
physical or operational changes are made to the plant. The proposed
change does not change how the plant would mitigate an accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are editorial and administrative in nature
and do not result in a change in the manner in which the plant
operates. The loss of function of any specific component will
continue to be addressed in its specific TS LCO and plant
configuration will be governed by the required actions of those
LCOs. The proposed changes are clarifications that do not degrade
the availability or capability of safety related equipment, and
therefore do not create the possibility of a new or different kind
of accident from any accident previously evaluated. There are no
design changes associated with the proposed changes, and the changes
do not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed). The changes do not
alter assumptions made in the safety analysis, and are consistent
with the safety analysis assumptions and current plant operating
practice. Due to the administrative nature of the changes, they
cannot be an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to TS 5.5.15 are clarifications and are
editorial and administrative in nature. No changes are made the LCOs
for plant equipment, the time required for the TS Required Actions
to be completed, or the out of service time for the components
involved. The proposed changes do not affect the safety analysis
acceptance criteria for any analyzed event, nor is there a change to
any safety analysis limit. The proposed changes do not alter the
manner in which safety limits, limiting safety system settings or
limiting conditions for operation are determined, nor is there any
adverse effect on those plant systems necessary to assure the
accomplishment of protection functions. The proposed changes will
not result in plant operation in a configuration outside the design
basis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
17: TSTF-284-A, Revision 3, ``Add `Met vs. Perform' to Technical
Specification 1.4, Frequency,'' TS 1.4, TS 3.4, TS 3.9, TS Pages 1.4-1,
1.4-4, 3.4.11-3, 3.4.12-4 and 3.9.4-2
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes insert a discussion paragraph into
Specification 1.4, and several new examples are added to facilitate
the use and application of SR Notes that utilize the terms ``met''
and ``perform.'' The changes also modify SRs in multiple
Specifications to appropriately use ``met'' and ``perform''
exceptions. The changes are administrative in nature because they
provide clarification and correction of existing expectations, and
therefore the proposed change does not increase the probability of
any accident previously evaluated. No physical or
[[Page 11486]]
operational changes are made to the plant. The proposed change does
not significantly change how the plant would mitigate an accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are administrative in nature and do not
result in a change in the manner in which the plant operates. The
proposed changes provide clarification and correction of existing
expectations that do not degrade the availability or capability of
safety related equipment, and therefore do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. There are no design changes associated with
the proposed changes, and the changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed). The changes do not alter assumptions made in the
safety analysis, and are consistent with the safety analysis
assumptions and current plant operating practice. Due to the
administrative nature of the changes, they cannot be an accident
initiator.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes are administrative in nature and do not
result in a change in the manner in which the plant operates. The
proposed changes provide clarification and correction of existing
expectations that do not degrade the availability or capability of
safety related equipment, or alter their operation. The proposed
changes do not affect the safety analysis acceptance criteria for
any analyzed event, nor is there a change to any safety analysis
limit. The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined, nor is there any adverse effect on those
plant systems necessary to assure the accomplishment of protection
functions. The proposed changes will not result in plant operation
in a configuration outside the design basis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
18: TSTF-308-A, Revision 1, ``Determination of Cumulative and Projected
Dose Contributions in RECP [Radioactive Effluent Controls Program],''
TS 5.5.4, TS Page 5.5-3
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Specification 5.5.4, ``Radioactive
Effluent Controls Program,'' paragraph e, to describe the original
intent of the dose projections. The cumulative and projection of
doses due to liquid releases are not an assumption in any accident
previously evaluated and have no effect on the mitigation of any
accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises Specification 5.5.4, ``Radioactive
Effluent Controls Program,'' paragraph e, to describe the original
intent of the dose projections. The cumulative and projection of
doses due to liquid releases are administrative tools to assure
compliance with regulatory limits. The proposed change revises the
requirement to clarify the intent, thereby improving the
administrative control over this process. As a result, any effect on
the margin of safety should be minimal.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
19: TSTF-312-A, Revision 1, ``Administrative Control of Containment
Penetrations,'' TS 3.9.4, TS Page 3.9.4-1
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would allow containment penetrations to be
unisolated under administrative controls during core alterations or
movement of irradiated fuel assemblies within containment. The
status of containment penetration flow paths (i.e., open or closed)
is not an initiator for any design basis accident or event, and
therefore the proposed change does not increase the probability of
any accident previously evaluated. The proposed change does not
affect the design of the primary containment, or alter plant
operating practices such that the probability of an accident
previously evaluated would be significantly increased. The proposed
change does not significantly change how the plant would mitigate an
accident previously evaluated, and is bounded by the fuel handling
accident (FHA) accident analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Allowing penetration flow paths to be open is not an initiator
for any accident. The proposed change to allow open penetration flow
paths will not affect plant safety functions or plant operating
practices such that a new or different accident could be created.
There are no design changes associated with the proposed changes,
and the change does not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed). The
change does not alter assumptions made in the safety analysis, and
is consistent with the safety analysis assumptions and current plant
operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
TS 3.9.4 provides measures to ensure that the dose consequences
of a postulated FHA inside containment are minimized. The proposed
change to LCO 3.9.4 will allow penetration flow path(s) to be open
during refueling operations under administrative control. These
administrative controls will can and will be achieved in the event
of an FHA inside containment, and will minimize dose consequences.
The proposed change is bounded by the existing FHA analysis. The
proposed change does not affect the safety analysis acceptance
criteria for any analyzed event, nor is there a change to any safety
analysis limit. The proposed change does not alter the manner in
which safety limits, limiting safety system settings or limiting
conditions for operation are determined, nor is there any adverse
effect on those plant systems necessary to assure the accomplishment
of protection functions. The proposed change will not result in
plant operation in a configuration outside the design basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
[[Page 11487]]
20: TSTF-314-A, Revision 0, ``Require Static and Transient
FQ Measurement,'' TS 3.1.4, 3.2.4, TS Pages 3.1.4-2, 3.2.4-
1, 3.2.4-3
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Required Actions of
Specification 3.1.4, ``Rod Group Alignment Limits,'' and
Specification 3.2.4, ``Quadrant Power Tilt Ratio,'' to require
measurement of both the steady state and transient portions of the
Heat Flux Hot Channel Factor, FQ(Z). This change will ensure that
the hot channel factors are within their limits when the rod
alignment limits or quadrant power tilt ratio are not within their
limits. The verification of hot channel factors is not an initiator
of any accident previously evaluated. The verification that both the
steady state and transient portion of FQ(Z) are within their limits
will ensure this initial assumption of the accident analysis is met
should a previously evaluated accident occur.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises the Required Actions in the
Specifications for Rod Group Alignment Limits and Quadrant Power
Tilt Ratio to require measurement of both the steady state and
transient portions of the Heat Flux Hot Channel Factor,
FQ(Z). This change is a correction that ensures that the
plant conditions are as assumed in the accident analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
21: TSTF-340-A, Revision 3, ``Allow 7 Day Completion Time for a
Turbine--Driven AFW Pump Inoperable,'' TS 3.7.5, TS Page 3.7.5-1
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Specification 3.7.5, ``Auxiliary
Feedwater (AFW) System,'' to allow a 7 day Completion Time to
restore an inoperable AFW turbine-driven pump in Mode 3 immediately
following a refueling outage, if Mode 2 has not been entered. An
inoperable AFW turbine-driven pump is not an initiator of any
accident previously evaluated. The ability of the plant to mitigate
an accident is no different while in the extended Completion Time
than during the existing Completion Time.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in of safety?
Response: No.
The proposed change revises Specification 3.7.5, ``Auxiliary
Feedwater (AFW) System,'' to allow a 7-day Completion Time to
restore an inoperable turbine-driven AFW pump in Mode 3 immediately
following a refueling outage if Mode 2 has not been entered. In Mode
3 immediately following a refueling outage, core decay heat is low
and the need for AFW is also diminished. The two operable motor
driven AFW pumps are available and there are alternate means of
decay heat removal if needed. As a result, the risk presented by the
extended Completion Time is minimal.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
22: TSTF-343-A, Revision 1, ``Containment Structural Integrity,'' TS
5.5, TS Page 5.5-16
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Technical Specifications (TS)
Administrative Controls programs for consistency with the
requirements of 10 CFR 50, paragraph 55a(g)(4) for components
classified as Code Class CC. The proposed changes affect the
frequency of visual examinations that will be performed for the
steel containment liner plate for the purpose of the Containment
Leakage Rate Testing Program.
The frequency of visual examinations of the containment and the
mode of operation during which those examinations are performed does
not affect the initiation of any accident previously evaluated. The
use of NRC approved methods and frequencies for performing the
inspections will ensure the containment continues to perform the
mitigating function assumed for accidents previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change revises the TS Administrative Controls
programs for consistency with the requirements of 10 CFR 50,
paragraph 55a(g)(4) for components classified as Code Class CC. The
proposed change affects the frequency of visual examinations that
will be performed for the steel containment liner plate for the
purpose of the Containment Leakage Rate Testing Program.
The proposed changes do not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed changes will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes revise the Technical Specifications (TS)
Administrative Controls programs for consistency with the
requirements of 10 CFR 50, paragraph 55a(g)(4) for components
classified as Code Class CC. The proposed change affects the
frequency of visual examinations that will be performed for the
steel containment liner plate for the purpose of the Containment
Leakage Rate Testing Program. The safety function of the containment
as a fission product barrier will be maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
23: TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 Allowing Shutdown
Cooling Loops Removal From Operation,'' TS 3.9.6, TS Page 3.9.6-1
1. Does the proposed amendment involve a significant increase in
the probability or
[[Page 11488]]
consequences of an accident previously evaluated?
Response: No.
The proposed change adds an LCO Note to LCO 3.9.6, ``RHR and
Coolant Circulation--Low Water Level,'' to allow securing the
operating train of Residual Heat Removal (RHR) for up to 15 minutes
to support switching operating trains. The allowance is restricted
to conditions in which core outlet temperature is maintained at
least 10 degrees F below the saturation temperature, when there are
no draining operations, and when operations that could reduce the
reactor coolant system (RCS) boron concentration are prohibited.
Securing an RHR train to facilitate the changing of the operating
train is not an initiator to any accident previously evaluated. The
restrictions on the use of the allowance ensure that an RHR train
will not be needed during the 15 minute period to mitigate any
accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change adds an LCO Note to LCO 3.9.6, ``RHR and
Coolant Circulation--Low Water Level,'' to allow securing the
operating train of RHR to support switching operating trains. The
allowance is restricted to conditions in which core outlet
temperature is maintained at least 10 degrees F below the saturation
temperature, when there are no draining operations, and when
operations that could reduce the reactor coolant system (RCS) boron
concentration are prohibited. With these restrictions, combined with
the short time frame allowed to swap operating RHR trains and the
ability to start an operating RHR train if needed, the occurrence of
an event that would require immediate operation of an RHR train is
extremely remote.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Leigh D. Perry, SVP & General Counsel of
Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness
Center Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J. Pascarelli.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power, Unit Nos. 1 and 2, Louisa County, Virginia
Date of amendment request: February 4, 2015. A publicly-available
version is in ADAMS under Accession No. ML15041A667.
Description of amendment request: The proposed license amendment
requests the changes to the Technical Specification (TS) TS 3.1.7, Rod
Position Indication, to provide an additional monitoring option for an
inoperable control rod position indicator. Specifically, the proposed
changes would allow monitoring of control rod drive mechanism
stationary gripper coil voltage every eight hours as an alternative to
using the movable in core detectors every eight hours to verify control
rod position.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change provides an alternative method for verifying
rod position of one rod. The proposed change meets the intent of the
current specification in that it ensures verification of position of
the rod once every 8 hours. The proposed change provides only an
alternative method of monitoring rod position and does not change
the assumptions or results of any previously evaluated accident.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change provides only an alternative method of
determining the position of one rod. No new accident initiators are
introduced by the proposed alternative manner of performing rod
position verification. The proposed change does not affect the
reactor protection system. Hence, no new failure modes are created
that would cause a new or different kind of accidents from any
accident previously evaluated.
Therefore, operation of the facility in accordance with the
proposed amendments would not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The basis of TS 3.1.7 states that the operability of the rod
position indicators 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 an alternative method for
determining the position of the affected rod. As a result, the
initial conditions of the accident analysis are preserved and the
consequences of previously analyzed accidents are unaffected.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
Based on the above, Dominion concludes that the proposed
amendment 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: Robert Pascarelli.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating
[[Page 11489]]
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 Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: May 29, 2013, as supplemented by
letters dated September 23, October 15, October 17, October 31, and
November 7, 2013, and January 7, March 13, April 29, and October 6,
2014, and January 15, 2015.
Brief description of amendment: The amendment revised the Renewed
Facility Operating License and associated Technical Specifications to
conform to the permanent shutdown and defueled status of the facility.
It also denied a proposal to delete paragraphs 1.B, 1.I, and 1.J of the
Kewaunee Operating License.
Date of issuance: February 13, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 215. A publicly-available version is in ADAMS under
Accession No. ML14237A045; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-43: The amendment
revised the renewed facility operating license and Technical
Specifications.
Date of initial notice in Federal Register: August 20, 2013 (78 FR
51224). The supplemental letters dated September 23, October 15,
October 17, October 31, and November 7, 2013, and January 7, March 13,
April 29, and October 6, 2014, and January 15, 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 Safety Evaluation dated February 13, 2015.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370 McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application for amendments: July 21, 2014.
Brief description of amendments: The amendment revises the licensed
operator training requirements to be consistent with the National
Academy for Nuclear Training (NANT) program. Additionally, the
amendment makes administrative changes to Technical Specification
Sections 5.1, ``Responsibility;'' 5.2, ``Organization;'' 5.3, ``Unit
Staff Qualifications;'' 5.5, ``Programs and Manuals;'' and for Catawba
and McGuire, Section 5.7, ``High Radiation Area.''
Date of issuance: February 12, 2015.
Effective date: This license amendment is effective as of its date
of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 273, 269, 276, 256, 389, 391, and 390. A publicly-
available version is available in ADAMS under Accession No.
ML15002A324.
Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-
17, DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2014 (79
FR 67199).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 12, 2015.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: December 17, 2012, as
supplemented by letters dated November 7, and December 4, 2013; January
6, May 22, June 30, August 7, September 24, and December 9, 2014.
Brief description of amendment: The amendment authorized the
transition of the Arkansas Nuclear One, Unit No. 2, fire protection
program to a risk-informed, performance-based program based on National
Fire Protection Association (NFPA) 805, in accordance with 10 CFR
50.48(c). NFPA 805 allows the use of performance-based methods such as
fire modeling and risk-informed methods such as fire probabilistic risk
assessment to demonstrate compliance with the nuclear safety
performance criteria.
Date of issuance: February 18, 2015.
Effective date: As of its date of issuance and shall be implemented
by 6 months from the date of issuance.
Amendment No.: 300. A publicly-available version is in ADAMS under
Accession No. ML14356A227; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPR-6: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: July 23, 2013 (78 FR
44171). The supplemental letters dated November 7 and December 4, 2013;
and January 6, May 22, June 30, August 7, September 24, and December 9,
2014, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 18, 2015.
No significant hazards consideration comments received: No.
Entergy Nuclear FitzPatrick, LLC and Entergy Nuclear Operations, Inc.,
Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego
County, New York
Date of amendment request: October 8, 2013, as supplemented by a
letter dated November 18, 2014.
Brief description of amendment: The amendment modifies the
Technical Specifications (TSs) to reduce the reactor steam dome
pressure associated
[[Page 11490]]
with the Reactor Core Safety Limit from 785 psig to 685 psig in TS
2.1.1.1 and TS 2.1.1.2. This change addresses the potential to not meet
the pressure/thermal power/minimal critical power ratio TS safety limit
during a pressure regulator failure-maximum demand (open) (PRFO)
transient. The PRFO transient was reported by General Electric as a
notification pursuant to Title 10 of the Code of Federal Regulations,
Part 21, ``Reporting of Defects and Noncompliance.''
Date of issuance: February 9, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 309. A publicly-available version is in ADAMS under
Accession No. ML15014A277; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-59: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 8, 2014 (79 FR
38589). The supplemental letter dated November 18, 2014, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 2015.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: November 14, 2013, as supplemented by
letters dated June 9, 2014, August 6, 2014, and October 9, 2014.
Description of amendment request: The amendment eliminates
operability requirements for secondary containment when handling
sufficiently decayed irradiated fuel or a fuel cask following a minimum
of 13 days after the permanent cessation of reactor operation.
Date of Issuance: February 12, 2015.
Effective date: The license amendment becomes effective 13 days
after the licensee's submittal of the certifications, as required by 10
CFR 50.82(a)(1)(i) and (ii).
Amendment No.: 262. A publicly-available version is in ADAMS under
Accession No. ML14304A588; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. DPR-28: The amendment revised the
Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 16, 2014 (79
FR 55511).
The supplemental letters dated June 9, 2014, August 6, 2014, and
October 9, 2014, 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 this amendment is contained
in a Safety Evaluation dated February 12, 2015.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: June 23, 2014.
Brief description of amendment: The amendment revised the Technical
Specification (TS) requirements to address NRC Generic Letter (GL)
2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay
Heat Removal, and Containment Spray Systems,'' as described in TSTF-
523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.''
Date of issuance: February 10, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 290. A publicly-available version is in ADAMS under
Accession No. ML15014A200; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-49: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: September 30, 2014 (79
FR 58820).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 2015.
No significant hazards consideration comments received: No
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham County, New Hampshire
Date of amendment request: June 24, 2014, as supplemented by letter
dated December 11, 2014.
Brief description of amendment: The amendment revised the Seabrook
Technical Specifications (TSs). Specifically, the amendment modifies
Seabrook TSs to address U.S. Nuclear Regulatory Commission Generic
Letter (GL) 2008-01, ``Managing Gas Accumulation in Emergency Core
Cooling, Decay Heat Removal, and Containment Spray Systems,'' as
described in TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing
Gas Accumulation.''
Date of issuance: February 6, 2015.
Effective date: As of its date of issuance and shall be implemented
within 60 days.
Amendment No.: 144. A publicly-available version is in ADAMS under
Accession No. ML14345A288; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-86: The amendment revised the
License and TS.
Date of initial notice in Federal Register: September 2, 2014 (79
FR 52066). The supplemental letter dated December 11, 2014, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 6, 2015.
No significant hazards consideration comments received: No.
South Carolina Electric and Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of amendment request: November 15, 2011, as supplemented by
letters dated November 22, 2011; January 26 and October 10, 2012;
February 1, April 1, October 14, and November 26, 2013; January 9,
February 25, May 2, May 11, August 14, October 9, and December 11,
2014.
Brief description of amendment: The amendment authorizes the
transition of the V.C. Summer fire protection program to a risk-
informed, performance-based program based on
[[Page 11491]]
National Fire Protection Association (NFPA) 805, ``Performance-Based
Standard for Fire Protection for Light Water Reactor Electric
Generating Plants, 2001 Edition'' (NFPA 805), in accordance with 10 CFR
50.48(c).
Date of issuance: February 11, 2015.
Effective date: This amendment is effective as of its date of
issuance and shall be implemented per the December 11, 2014,
supplement, Attachment S, Table S-2 ``Implementation Items'', requiring
full implementation by March 31, 2016.
Amendment No.: 199. A publicly-available version is in ADAMS under
Accession No. ML14287A289; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-12: Amendment revised
the Facility Operating License.
Date of initial notice in Federal Register: August 14, 2012 (77 FR
48561). The supplemental letters dated November 22, 2011; October 10,
2012; February 1, April 1, October 14, and November 26, 2013; January
9, February 25, May 2, May 11, August 14, October 9, and December 11,
2014, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 11, 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. 50-366, Edwin I. Hatch Nuclear
Plant (HNP), Unit No. 2, Appling County, Georgia
Date of amendment request: August 8, 2014, as supplemented by
letters dated September 8 and October 24, 2014.
Brief description of amendments: The amendment revises the
Technical Specification value of the Safety Limit Minimum Critical
Power Ratio to support operation in the next fuel cycle.
Date of issuance: February 18, 2015.
Effective date: As of the date of issuance and shall be implemented
prior to reactor startup following the HNP, Unit 2, spring 2015
refueling outage.
Amendment No(s).: 218. A publicly-available version is in ADAMS
under Accession No. ML15020A434; documents related to this amendment
are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendment
revised the licenses and the Technical Specifications.
Date of initial notice in Federal Register: January 6, 2015, (80 FR
536).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 18, 2015.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project (STP), Units 1 and 2, Matagorda County, Texas
Date of amendment request: July 23, 2013, as supplemented by
letters dated May 12 (two letters), May 19, and December 17, 2014.
Brief description of amendments: The amendments revised the STP,
Units 1 and 2, Fire Protection Program (FPP) related to the alternate
shutdown capability. Specifically, it approves the following operator
actions in the control room prior to evacuation due to a fire for
meeting the alternate shutdown capability, in addition to manually
tripping the reactor that is currently credited in the STP, Units 1 and
2, FPP licensing basis:
Initiate main steam line isolation
Closing the pressurizer power-operated relief valves block
valves
Securing all reactor coolant pumps
Closing feedwater isolation valves
Securing the startup feedwater pump
Isolating reactor coolant system letdown
Securing the centrifugal charging pumps
In addition, the licensee credits the automatic trip of the main
turbine upon the initiation of a manual reactor trip for meeting the
alternate shutdown capability.
Date of issuance: February 13, 2015.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: Unit 1--203; Unit 2--191. A publicly-available
version is in ADAMS under Accession No. ML14339A170; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: October 29, 2013 (78 FR
64546). The supplements dated May 12 (two letters), May 19, and
December 17, 2014, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 13, 2015.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: December 18, 2013, as supplemented by
letter dated June 13, 2014.
Brief description of amendment: The amendment revised the Technical
Specification (TS) 3.4.9, ``RCS [Reactor Coolant System] Pressure and
Temperature (P/T) Limits,'' Figures 3.4.9-1 through 3.4.9-2. The P/T
limits are based on proprietary topical report NEDC-33178P-A, Revision
1, ``GE [General Electric] Hitachi Nuclear Energy Methodology for
Development of Reactor Pressure Vessel Pressure-Temperature Curves.''
NEDO-33178-A, Revision 1 is the non-proprietary version of the NRC-
approved topical report.
Date of issuance: February 2, 2015.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 287. A publicly available version is in ADAMS under
Accession No. ML14325A501; documents related to this amendment are
listed in the Safety Evaluation (SE) enclosed with the amendment.
Renewed Facility Operating License No. DPR-33: Amendment revised
the TSs and the Operating License.
Date of initial notice in Federal Register: May 6, 2014 (79 FR
25902). The supplemental letter dated June 13, 2014, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in the SE dated February 2, 2015.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 23rd day of February 2015.
[[Page 11492]]
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2015-04298 Filed 3-2-15; 8:45 am]
BILLING CODE 7590-01-P