80 FR 11472 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Federal Register Volume 80, Issue 41 (March 3, 2015)

Page Range		11472-11492
FR Document		2015-04298

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.

Federal Register, Volume 80 Issue 41 (Tuesday, March 3, 2015)

[Federal Register Volume 80, Number 41 (Tuesday, March 3, 2015)]
[Notices]
[Pages 11472-11492]
From the Federal Register Online [www.thefederalregister.org]
[FR Doc No: 2015-04298]

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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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission
(NRC) is publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from 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 http://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 http://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 http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to 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 http://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 http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends

[[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 http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's public
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to
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
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)-(iii).
For further details with respect to 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.

[[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.

[[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.

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.

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.

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

11474 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

to rely to establish those facts or expert days prior to the filing deadline, the system. To be timely, an electronic
opinion. The petition must include participant should contact the Office of filing must be submitted to the E-Filing
sufficient information to show that a the Secretary by email at system no later than 11:59 p.m. Eastern
genuine dispute exists with the hearing.docket@nrc.gov, or by telephone Time on the due date. Upon receipt of
applicant on a material issue of law or at 301–415–1677, to request (1) a digital a transmission, the E-Filing system
fact. Contentions shall be limited to identification (ID) certificate, which time-stamps the document and sends
matters within the scope of the allows the participant (or its counsel or the submitter an email notice
amendment under consideration. The representative) to digitally sign confirming receipt of the document. The
contention must be one which, if documents and access the E-Submittal E-Filing system also distributes an email
proven, would entitle the requestor/ server for any proceeding in which it is notice that provides access to the
petitioner to relief. A requestor/ participating; and (2) advise the document to the NRC’s Office of the
petitioner who fails to satisfy these Secretary that the participant will be General Counsel and any others who
requirements with respect to at least one submitting a request or petition for have advised the Office of the Secretary
contention will not be permitted to hearing (even in instances in which the that they wish to participate in the
participate as a party. participant, or its counsel or proceeding, so that the filer need not
Those permitted to intervene become representative, already holds an NRC- serve the documents on those
parties to the proceeding, subject to any issued digital ID certificate). Based upon participants separately. Therefore,
limitations in the order granting leave to this information, the Secretary will applicants and other participants (or
intervene, and have the opportunity to establish an electronic docket for the their counsel or representative) must
participate fully in the conduct of the hearing in this proceeding if the apply for and receive a digital ID
hearing. Secretary has not already established an certificate before a hearing request/
If a hearing is requested, the electronic docket. petition to intervene is filed so that they
Commission will make a final Information about applying for a can obtain access to the document via
determination on the issue of no digital ID certificate is available on the the E-Filing system.
significant hazards consideration. The NRC’s public Web site at http:// A person filing electronically using
final determination will serve to decide www.nrc.gov/site-help/e-submittals/ the NRC’s adjudicatory E-Filing system
when the hearing is held. If the final getting-started.html. System may seek assistance by contacting the
determination is that the amendment requirements for accessing the E- NRC Meta System Help Desk through
request involves no significant hazards Submittal server are detailed in the the ‘‘Contact Us’’ link located on the
consideration, the Commission may NRC’s ‘‘Guidance for Electronic NRC’s public Web site at http://
issue the amendment and make it Submission,’’ which is available on the www.nrc.gov/site-help/e-
immediately effective, notwithstanding agency’s public Web site at http:// submittals.html, by email to
the request for a hearing. Any hearing www.nrc.gov/site-help/e- MSHD.Resource@nrc.gov, or by a toll-
held would take place after issuance of submittals.html. Participants may free call at 1–866–672–7640. The NRC
the amendment. If the final attempt to use other software not listed Meta System Help Desk is available
determination is that the amendment on the Web site, but should note that the between 8 a.m. and 8 p.m., Eastern
request involves a significant hazards NRC’s E-Filing system does not support Time, Monday through Friday,
consideration, then any hearing held unlisted software, and the NRC Meta excluding government holidays.
would take place before the issuance of System Help Desk will not be able to Participants who believe that they
any amendment unless the Commission offer assistance in using unlisted have a good cause for not submitting
finds an imminent danger to the health software. documents electronically must file an
or safety of the public, in which case it If a participant is electronically exemption request, in accordance with
will issue an appropriate order or rule submitting a document to the NRC in 10 CFR 2.302(g), with their initial paper
under 10 CFR part 2. accordance with the E-Filing rule, the filing requesting authorization to
participant must file the document continue to submit documents in paper
B. Electronic Submissions (E-Filing) using the NRC’s online, Web-based format. Such filings must be submitted
All documents filed in NRC submission form. In order to serve by: (1) First class mail addressed to the
adjudicatory proceedings, including a documents through the Electronic Office of the Secretary of the
request for hearing, a petition for leave Information Exchange System, users Commission, U.S. Nuclear Regulatory
to intervene, any motion or other will be required to install a Web Commission, Washington, DC 20555–
document filed in the proceeding prior browser plug-in from the NRC’s Web 0001, Attention: Rulemaking and
to the submission of a request for site. Further information on the Web- Adjudications Staff; or (2) courier,
hearing or petition to intervene, and based submission form, including the express mail, or expedited delivery
documents filed by interested installation of the Web browser plug-in, service to the Office of the Secretary,
governmental entities participating is available on the NRC’s public Web Sixteenth Floor, One White Flint North,
under 10 CFR 2.315(c), must be filed in site at http://www.nrc.gov/site-help/e- 11555 Rockville Pike, Rockville,
accordance with the NRC’s E-Filing rule submittals.html. Maryland 20852, Attention: Rulemaking
(72 FR 49139; August 28, 2007). The E- Once a participant has obtained a and Adjudications Staff. Participants
Filing process requires participants to digital ID certificate and a docket has filing a document in this manner are
submit and serve all adjudicatory been created, the participant can then responsible for serving the document on
documents over the internet, or in some submit a request for hearing or petition all other participants. Filing is
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cases to mail copies on electronic for leave to intervene. Submissions considered complete by first-class mail
storage media. Participants may not should be in Portable Document Format as of the time of deposit in the mail, or
submit paper copies of their filings (PDF) in accordance with NRC guidance by courier, express mail, or expedited
unless they seek an exemption in available on the NRC’s public Web site delivery service upon depositing the
accordance with the procedures at http://www.nrc.gov/site-help/e- document with the provider of the
described below. submittals.html. A filing is considered service. A presiding officer, having
To comply with the procedural complete at the time the documents are granted an exemption request from
requirements of E-Filing, at least ten 10 submitted through the NRC’s E-Filing using E-Filing, may require a participant

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11476 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

consideration, which is presented the Technical Specifications. This is a The proposed TS 3.9.6 (Refuel
below: requirement as outlined in [10 CFR Machine) and TS 3.9.7 (Crane Travel)
1. Does the proposed change involve 50.36]. relocation to the Waterford 3 TRM is
a significant increase in the probability [10 CFR 50.36(b)] states the technical administrative in nature because all
or consequences of an accident specifications will be derived from the requirements will be relocated. Any
previously evaluated? analyses and evaluation included in the changes after being relocated to the
Response: No. safety analysis report. Waterford 3 TRM will require that the
This proposed change relocates [10 CFR 50.36(c)(2)(i)] states that [10 CFR 50.59] process be entered
Technical Specifications (TS) 3.9.6 [‘‘]the limiting conditions for operation ensuring the public health and safety is
(Refuel Machine) and TS 3.9.7 (Crane are the lowest functional capability or maintained. By using the [10 CFR 50.59]
Travel) to the Waterford 3 Technical performance levels of equipment process for future changes, the
Requirements Manual (TRM). This is required for safe operation of the regulatory requirements ensure that no
consistent with the requirements of [10 facility[. . . .’’] [10 CFR 50.36(c)(2)(ii)] significant reduction in the margin of
CFR 50.36(c)(2)(ii)] and aligns with provides the four criteria in which any safety occurs.
NUREG–1432 (Combustion Engineering one met requires a limiting condition for In addition, the TS 3.9.6 and TS 3.9.7
Standard Technical Specifications). operation. The proposed change requirements do not prevent the design
The applicable TS 3.9.6 and TS 3.9.7 basis accident conditions from
demonstrated that the [10 CFR
design basis accident is the Fuel occurring and do not limit the severity
50.36(c)(2)(ii)] criteria were not met and
Handling Accident (FHA) described in of the accident. Thus, TS 3.9.6 and TS
the relocation to the TRM is allowable.
the Updated Final Safety Analysis 3.9.7 relocation will not adversely
By not meeting the [10 CFR
Report (UFSAR) Section 15.7.3.4. The impact the accident analyses and will
50.36(c)(2)(ii)] criteria for inclusion into
limiting FHA results in all the fuel pins not cause a significant reduction in the
the TS means that TS 3.9.6 and TS 3.9.7
in the dropped and impacted fuel margin of safety.
do not impact the accident analyses
assemblies failing (472 pins or 236 per
previously evaluated and would not The NRC staff has reviewed the
assembly). The analysis assumes that a
create the possibility of a new or licensee’s analysis and, based on this
fuel assembly is dropped as an initial
different kind of accident. review, it appears that the three
condition and no equipment or
Specifically, TS 3.9.6 and TS 3.9.7 standards of 10 CFR 50.92(c) are
intervention can prevent the initiating
equipment are not instrumentation used satisfied. Therefore, the NRC staff
condition. The proposed change was
to detect, and indicate in the control proposes to determine that the
evaluated against [10 CFR
room, a significant abnormal amendment request involves no
50.36(c)(2)(ii)] criteria and shows no
degradation of the reactor coolant significant hazards consideration.
impact to the lowest functional
pressure boundary (Criterion 1). TS Attorney for licensee: Joseph A.
capability or performance levels of
3.9.6 and TS 3.9.7 do not contain a Aluise, Associate General Council—
equipment required for safe operation of
process variable, design feature, or Nuclear, Entergy Services, Inc., 639
the facility because the TS 3.9.6 and TS
operating restriction that is an initial Loyola Avenue, New Orleans, Louisiana
3.9.7 requirements do not prevent the
condition of a Design Basis Accident or 70113.
accident conditions from occurring and
Transient analysis that either assumes NRC Branch Chief: Meena K. Khanna.
do not limit the severity of the accident.
the failure of or presents a challenge to
Since, the dropped fuel assembly and Exelon Generation Company, LLC
the integrity of a fission product barrier
the impacted fuel assembly are both (EGC), Docket No. 50–410, Nine Mile
(Criterion 2). TS 3.9.6 and TS 3.9.7 does
already failed in the design basis Point Nuclear Station, Unit 2 (NMP2),
not contain a structure, system, or
accident scenario, this change could not Oswego County, New York
component that is part of the primary
result in a significant increase in the
success path and which functions or Date of amendment request:
accident consequences. The TS 3.9.6
actuates to mitigate a Design Basis November 17, 2014. A publicly
and TS 3.9.7 equipment are not required
Accident or Transient that either available version is in ADAMS under
to respond, mitigate, or terminate any
assumes the failure of or presents a Accession No. ML14321A744.
design basis accident, thus this change
challenge to the integrity of a fission Description of amendment request:
will not adversely impact the likelihood
product barrier (Criterion 3). Lastly, TS The proposed amendment would revise
or probability of a design basis accident.
The TS 3.9.6 and TS 3.9.7 3.9.6 and TS 3.9.7 do not contain a the NMP2 Technical Specification (TS)
requirements do not prevent the structure, system, or component which Allowable Value for the Main Steam
accident conditions from occurring and operating experience or probabilistic Line Tunnel Lead Enclosure
do not limit the severity of the accident. safety assessment has shown to be Temperature-High instrumentation from
Therefore the TS 3.9.6 and TS 3.9.7 significant to public health and safety an ambient temperature dependent
relocation to the TRM would not cause (Criterion 4). (variable setpoint) to ambient
a significant increase in the accident TS 3.9.6 and 3.9.7 are not required to temperature independent (constant
probability or accident consequences. meet the lowest functional capability or Allowable Value). The changes would
2. Does the proposed change create performance levels of equipment delete Surveillance Requirement (SR)
the possibility of a new or different kind required for safe operation of the 3.3.6.1.2 and revise the Allowable Value
of accident from any accident facility. for Function 1.g on Table 3.3.6.1–1,
previously evaluated? Therefore, the accident analyses are ‘‘Primary Containment Isolation
not impacted and the possibility of a
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Response: No. Instrumentation.’’
This proposed change relocates TS new or different kind of accident from Basis for proposed no significant
3.9.6 (Refuel Machine) and TS 3.9.7 any accident previously evaluated has hazards consideration determination:
(Crane Travel) to the Waterford 3 TRM. not changed. As required by 10 CFR 50.91(a), the
In general, Technical Specifications are 3. Does the proposed change involve licensee has provided its analysis of the
based upon the accident analyses. The a significant reduction in a margin of issue of no significant hazards
accident analyses assumptions and safety? consideration, which is presented
initial conditions must be protected by Response: No. below:

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Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11477

1. Does the proposed amendment safety because the changes eliminate the ‘‘Risk-Informed Technical
involve a significant increase in the temperature setpoint dependency on Specifications Initiative 5b, Risk-
probability or consequences of an lead enclosure temperature while Informed Method for Control of
accident previously evaluated? maintaining the existing upper AV Surveillance Frequencies’’ (ADAMS
Response: No. [Allowable Value] = 175.6 °F, that was Accession No. ML071360456). The
The proposed changes do not involve previously evaluated and approved. licensee stated that the NEI 04–10
a significant increase in the probability There is no adverse impact on the methodology provides reasonable
or consequences of an accident existing equipment capability as well as acceptance guidelines and methods for
previously evaluated because the associated structures. The increase in evaluating the risk increase of proposed
performance of any equipment credited the steam leak rate and associated crack changes to surveillance frequencies,
in the radiological consequences of an size continues to be well below the leak consistent with Regulatory Guide 1.177,
accident is not affected by the change in rate associated with critical crack size ‘‘An Approach for Plant-Specific, Risk-
the leak detection capability. that leads to pipe break. The proposed Informed Decisionmaking: Technical
The Main Steam Line Tunnel Lead changes continue to provide the same Specifications’’ (ADAMS Accession No.
Enclosure Temperature—High is level of protection against a main steam ML003740176). The licensee stated that
provided to detect a steam leak in the line break as the existing setpoint the changes are consistent with NRC-
lead enclosure and provides diversity to values. approved Technical Specification Task
the high flow instrumentation. This Therefore, the proposed changes do Force (TSTF) Improved Standard
function provides a mitigating action for not involve a significant reduction in a Technical Specifications Change
a steam leak in the Main Steam Line margin of safety. Traveler TSTF–425, ‘‘Relocate
Tunnel Lead Enclosure, which could Surveillance Frequencies to Licensee
lead to a pipe break. This function does The NRC staff has reviewed the
licensee’s analysis and, based on this Control—RITSTF [Risk-Informed
not affect any accident precursors, and Technical Specifications Task Force]
the proposed changes do not affect the review, it appears that the three
standards of 10 CFR 50.92(c) are Initiative 5b,’’ Revision 3 (ADAMS
leak detection capability. Additionally, Accession No. ML090850642). The
the proposed changes do not degrade satisfied. Therefore, the NRC staff
proposes to determine that the Federal Register notice published on
the performance of or increase the July 6, 2009 (74 FR 31996), announced
challenges to any safety systems amendment request involves no
significant hazards consideration. the availability of TSTF–425, Revision
assumed to function in the accident 3. In the supplement dated January 28,
analysis. Attorney for licensee: J. Bradley
Fewell, Senior Vice President, 2015, the licensee requested (1)
Therefore, the proposed changes do additional surveillance frequencies be
not involve a significant increase in the Regulatory Affairs, Nuclear, and General
Counsel, Exelon Generation Company, relocated to the licensee-controlled
probability or consequences of an program, (2) editorial changes, (3)
accident previously evaluated. LLC, 4300 Winfield Road, Warrenville,
IL 60555. administrative deviations from TSTF–
2. Does the proposed amendment 425, and (4) other changes resulting
create the possibility of a new or NRC Branch Chief: Benjamin G.
Beasley. from differences between the St. Lucie
different kind of accident from any Plant TSs and the TSs on which TSTF–
accident previously evaluated? Florida Power and Light Company, et al. 425 was based.
Response: No. (FPL), Docket Nos. 50–335 and 50–389, Basis for proposed no significant
The proposed changes do not create St. Lucie Plant, Unit Nos. 1 and 2, St. hazards consideration determination:
the possibility of a new or different kind Lucie County, Florida As required by 10 CFR 50.91(a), the
of accident from any accident licensee provided its analysis of the
previously evaluated because the Date of amendment request: February
20, 2014, as supplemented by letters issue of no significant hazards
proposed changes do not add or remove consideration, which is presented
equipment and do not physically alter dated December 11, 2014, January 13
and January 28, 2015. Publicly-available below:
the isolation instrumentation. In
addition, the Main Steam Line Tunnel in ADAMS under Accession Nos. 1. Does the proposed change involve a
Lead Enclosure LDS [Leak Detection ML14070A087, ML14349A333, significant increase in the probability or
ML15029A497 and ML15042A122. consequences of an accident previously
System] is not utilized in a different evaluated?
manner. The proposed changes do not Description of amendment request:
Response: No.
introduce any new accident initiators The NRC staff has previously made a The proposed change relocates the
and new failure modes, nor do they proposed determination that the specified frequencies for periodic
reduce or adversely affect the amendment request dated February 20, surveillance requirements to licensee control
capabilities of any plant structure, 2014, involves no significant hazards under a new Surveillance Frequency Control
system, or component to perform their consideration (see 79 FR 42550, July 22, Program. Surveillance frequencies are not an
2014). Subsequently, by letter dated initiator to any accident previously
safety function. The Main Steam Line evaluated. As a result, the probability of any
Tunnel Lead Enclosure LDS will January 28, 2015, the licensee provided
accident previously evaluated is not
continue to be operated in the same additional information that expanded significantly increased. The systems and
manner. the scope of the amendment request as components required by the technical
Therefore, the proposed changes do originally noticed. Accordingly, this specifications for which the surveillance
not create the possibility of a new or notice supersedes the previous notice in frequencies are relocated are still required to
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different kind of accident from any its entirety. be operable, meet the acceptance criteria for
accident previously evaluated. The amendment would revise the the surveillance requirements, and be
Technical Specifications (TSs) by capable of performing any mitigation
3. Does the proposed amendment
relocating specific surveillance function assumed in the accident analysis.
involve a significant reduction in a As a result, the consequences of any accident
margin of safety? frequency requirements to a licensee- previously evaluated are not significantly
Response: No. controlled program with increased.
The proposed changes do not involve implementation of Nuclear Energy Therefore, the proposed change does not
a significant reduction in a margin of Institute (NEI) 04–10 (Revision 1), involve a significant increase in the

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11478 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

probability or consequences of any accident Description of amendment request: limiting single failures will be
previously evaluated. The amendment would revise Technical introduced as a result of the proposed
2. Does the proposed amendment create Specification (TS) 3/4.5.2, ‘‘ECCS change. There will be no adverse effect
the possibility of a new or different kind of [Emergency Core Cooling System] or challenges imposed on any SSC as a
accident from any previously evaluated?
Response: No. Subsystems—Tavg [average temperature] result of the proposed change.
No new or different accidents result from Greater Than or Equal to 350 °F [degrees Therefore, the proposed changes do
utilizing the proposed change. The changes Fahrenheit],’’ to correct non- not create the possibility of a new or
do not involve a physical alteration of the conservative TS requirements. The different kind of accident from any
plant (i.e., no new or different type of licensee also requested editorial changes previously evaluated.
equipment will be installed) or a change in to the TS. 3. Does the proposed change involve
the methods governing normal plant Basis for proposed no significant a significant reduction in the margin of
operation. In addition, the changes do not hazards consideration determination: safety?
impose any new or different requirements. As required by 10 CFR 50.91(a), the No. Margin of safety is related to
The changes do not alter assumptions made confidence in the ability of the fission
in the safety analysis assumptions and
licensee provided its analysis of the
current plant operating practice. issue of no significant hazards product barriers to perform their
Therefore, the proposed changes do not consideration, which is presented as accident mitigation functions. The
create the possibility of a new or different follows: proposed TS changes involve TS 3.5.2
kind of accident from any accident 1. Does the proposed change involve Action ‘a’, new TS 3.5.2 Action ‘h’, and
previously evaluated. a significant increase in the probability the provision in SR 4.5.2.a to address
3. Does the proposed change involve a or consequences of an accident non-conservative TS requirements.
significant reduction in the margin of safety? previously evaluated? Editorial changes are also proposed for
Response: No. No. The proposed TS changes involve consistency and clarity. The proposed
The design, operation, testing methods, TS 3.5.2 Action ‘a’, new TS 3.5.2 Action changes provide better assurance that
and acceptance criteria for systems, ‘h’, and the provision in SR
structures, and components (SSCs), specified the ECCS systems, subsystems, and
in applicable codes and standards (or
[Surveillance Requirement] 4.5.2.a to components are properly aligned to
alternatives approved for use by the NRC) address non-conservative TS support safe reactor operation consistent
will continue to be met as described in the requirements. Editorial changes are also with the licensing basis requirements.
plant licensing basis (including the final proposed for consistency and clarity. The proposed changes do not physically
safety analysis report and bases to TS), since These changes do not affect any alter any SSC. There will be no effect on
these are not affected by changes to the precursors to any accident previously those SSCs necessary to assure the
surveillance frequencies. Similarly, there is evaluated and subsequently, will not accomplishment of specified functions.
no impact to safety analysis acceptance impact the probability or consequences There will be no impact on the
criteria as described in the plant licensing of an accident previously evaluated.
basis. To evaluate a change in the relocated overpower limit, departure from
Furthermore, these changes do not nucleate boiling ratio (DNBR) limits,
surveillance frequency, FPL will perform a
probabilistic risk evaluation using the adversely affect mitigation equipment or loss of cooling accident peak cladding
guidance contained in NRC-approved NEI strategies. temperature (LOCA PCT), or any other
04–10, Revision 1 in accordance with the TS Therefore, the proposed changes do margin of safety. The applicable
Surveillance Frequency Control Program. NEI not involve a significant increase in the radiological dose consequence
04–10, Revision 1, methodology provides probability or consequences of an acceptance criteria will continue to be
reasonable acceptance guidelines and accident previously evaluated. met. Therefore, the proposed changes do
methods for evaluating the risk increase of 2. Does the proposed change create not involve a significant reduction in a
proposed changes to surveillance frequencies the possibility of a new or different kind
consistent with Regulatory Guide (RG) 1.177. margin of safety.
of accident from any previously The NRC staff has reviewed the
Therefore, the proposed changes do evaluated? licensee’s analysis and, based on this
not involve a significant reduction in a No. The proposed TS changes involve review, it appears that the three
margin of safety. TS 3.5.2 Action ‘a’, new TS 3.5.2 Action standards of 10 CFR 50.92(c) are
The NRC staff has reviewed the ‘h’, and the provision in SR 4.5.2.a to satisfied. Therefore, the NRC staff
licensee’s analysis and, based on this address non-conservative TS proposes to determine that the
review, it appears that the three requirements. Editorial changes are also amendment request involves no
standards of 10 CFR 50.92(c) are proposed for consistency and clarity. significant hazards consideration.
satisfied. Therefore, the NRC staff The proposed changes provide better Attorney for licensee: William S.
proposes to determine that the assurance that the ECCS systems, Blair, Managing Attorney—Nuclear,
amendment request involves no subsystems, and components are Florida Power & Light Company, P.O.
significant hazards consideration. properly aligned to support safe reactor Box 14000, Juno Beach, FL 33408–0420.
Attorney for licensee: William S. operation consistent with the licensing NRC Branch Chief: Shana R. Helton.
Blair, Managing Attorney—Nuclear, basis requirements. The proposed
changes do not introduce new modes of Indiana Michigan Power Company
Florida Power & Light Company, 700
plant operation and do not involve (I&M), Docket Nos. 50–315 and 50–316,
Universe Boulevard, MS LAW/JB, Juno
physical modifications to the plant (no Donald C. Cook Nuclear Plant, Units 1
Beach, FL 33408–0420.
NRC Branch Chief: Shana R. Helton. new or different type of equipment will and 2, Berrien County, Michigan
be installed). There are no changes in Date of amendment request: February
Florida Power and Light Company,
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the method by which any safety related 6, 2015. A publicly-available version is
Docket Nos. 50–250 and 50–251, Turkey plant structure, system, or component in ADAMS under Accession No.
Point Nuclear Generating, Unit Nos. 3 (SSC) performs its specified safety ML15041A069.
and 4, Miami-Dade County, Florida function. As such, the plant conditions Description of amendment request:
Date of amendment request: for which the design basis accident The proposed amendments would
November 13, 2014. A publicly- analyses were performed remain valid. modify the technical specifications
available version is in ADAMS under No new accident scenarios, transient requirements for unavailable barriers by
Accession No. ML14337A013. precursors, failure mechanisms, or adding limiting condition for operation

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Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11479

(LCO) 3.0.8. The changes are consistent lead to an accident whose consequences Frequency Control Program (SFCP), with
with the NRC approved Technical exceed the consequences of accidents implementation of Nuclear Energy Institute
Specification Task Force (TSTF) previously evaluated. The addition of a (NEI) 04–10, ‘‘Risk-Informed Technical
Specifications Initiative 5b, Risk-Informed
Standard Technical Specification requirement to assess and manage the Method for Control of Surveillance
change TSTF–427, ‘‘Allowance for Non- risk introduced by this change will Frequencies’’ (ADAMS Accession No.
Technical Specification Barrier further minimize possible concerns. ML071360456). The changes are consistent
Degradation on Supported System Thus, this change does not create the with NRC-approved TS Task Force (TSTF)
OPERABILITY,’’ Revision 2. possibility of a new or different kind of Standard TS change TSTF–425, ‘‘Relocate
Basis for proposed no significant accident from an accident previously Surveillance Frequencies to Licensee
hazards consideration determination: evaluated. Control-Risk Informed Technical
As required by 10 CFR 50.91(a), the Specifications Task Force (RITSTF) Initiative
licensee has affirmed the applicability Criterion 3—The Proposed Change Does 5b,’’ Revision 3 (ADAMS Accession No.
of the model proposed no significant Not Involve a Significant Reduction in ML090850642). The Federal Register notice
a Margin of Safety published on July 6, 2009 (74 FR 31996),
hazards consideration published on announced the availability of this TSTF
October 3, 2006 (71 FR 58444), ‘‘Notice The proposed change allows a delay improvement, and included a model no
of Availability of the Model Safety time for entering a supported system TS significant hazards consideration and safety
Evaluation.’’ The findings presented in when the inoperability is due solely to evaluation.
that evaluation are presented below: an unavailable barrier, if risk is assessed Basis for proposed no significant hazards
and managed. The postulated initiating consideration determination: An analysis of
Criterion 1—The Proposed Change Does the no significant hazards consideration was
Not Involve a Significant Increase in the events which may require a functional
barrier are limited to those with low presented in the TSTF–425. The licensee has
Probability or Consequences of an affirmed its applicability of the model no
Accident Previously Evaluated frequencies of occurrence, and the significant hazards consideration, which is
overall TS system safety function would presented below:
The proposed change allows a delay still be available for the majority of
time for entering a supported system 1. Does the proposed change involve
anticipated challenges. The risk impact a significant increase in the probability
technical specification (TS) when the of the proposed TS changes was
inoperability is due solely to an or consequences of any accident
assessed following the three-tiered previously evaluated?
unavailable barrier if risk is assessed approach recommended in RG
and managed. The postulated initiating Response: No.
[Regulatory Guide] 1.177. A bounding The proposed change relocates the
events which may require a functional risk assessment was performed to justify specified frequencies for periodic
barrier are limited to those with low the proposed TS changes. This surveillance requirements to licensee control
frequencies of occurrence, and the application of LCO 3.0.8 is predicated under a new Surveillance Frequency Control
overall TS system safety function would upon the licensee’s performance of a Program. Surveillance frequencies are not an
still be available for the majority of risk assessment and the management of initiator to any accident previously
anticipated challenges. Therefore, the plant risk. The net change to the margin evaluated. As a result, the probability of any
probability of an accident previously accident previously evaluated is not
of safety is insignificant as indicated by significantly increased. The systems and
evaluated is not significantly increased, the anticipated low levels of associated
if at all. The consequences of an components required by the technical
risk (ICCDP [incremental conditional specifications for which the surveillance
accident while relying on the allowance core damage probability] and ICLERP frequencies are relocated are still required to
provided by proposed LCO 3.0.8 are no [incremental large early release be operable, meet the acceptance criteria for
different than the consequences of an probability]) as shown in Table 1 of the surveillance requirements, and be
accident while relying on the TS Section 3.1.1 in the Safety Evaluation. capable of performing any mitigation
required actions in effect without the Therefore, this change does not involve function assumed in the accident analysis.
allowance provided by proposed LCO As a result, the consequences of any accident
a significant reduction in a margin of previously evaluated are not significantly
3.0.8. Therefore, the consequences of an safety.
accident previously evaluated are not increased.
The NRC staff has reviewed the analysis Therefore, the proposed change does not
significantly affected by this change. involve a significant increase in the
and, based on this review, it appears that the
The addition of a requirement to assess probability or consequences of an accident
three standards of 10 CFR 50.92(c) are
and manage the risk introduced by this satisfied. Therefore, the NRC staff proposes to previously evaluated.
change will further minimize possible determine that the amendment requests 2. Does the proposed change create the
concerns. Therefore, this change does involve no significant hazards consideration. possibility of a new or different kind of
not involve a significant increase in the Attorney for licensee: Robert B. Haemer, accident from any accident previously
probability or consequences of an Senior Nuclear Counsel, One Cook Place, evaluated?
accident previously evaluated. Bridgman, Michigan 49106. Response: No.
NRC Branch Chief: David L. Pelton. No new or different accidents result from
Criterion 2—The Proposed Change Does PPL Susquehanna, LLC, Docket Nos. 50–387
utilizing the proposed change. The changes
Not Create the Possibility of a New or do not involve a physical alteration of the
and 50–388, Susquehanna Steam Electric
Different Kind of Accident From any plant (i.e., no new or different type of
Station, Units 1 and 2, Luzerne County,
Previously Evaluated equipment will be installed) or a change in
Pennsylvania
the methods governing normal plant
The proposed change does not Date of amendment requests: October 27, operation. In addition, the changes do not
involve a physical alteration of the plant 2014. A publicly-available version is impose any new or different requirements.
available in ADAMS under Accession No. The changes do not alter assumptions made
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(no new or different type of equipment
will be installed). Allowing delay times ML14317A052. in the safety analysis. The proposed changes
Description of amendment requests: The are consistent with the safety analysis
for entering supported system TS when
proposed amendments will modify the assumptions and current plant operating
inoperability is due solely to an Susquehanna technical specifications (TS). practice.
unavailable barrier, if risk is assessed Specifically, the proposed amendments will Therefore, the proposed changes do not
and managed, will not introduce new modify the TS by relocating specific create the possibility of a new or different
failure modes or effects and will not, in surveillance frequencies to a licensee- kind of accident from any accident
the absence of other unrelated failures, controlled program, the Surveillance previously evaluated.

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11480 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

3. Does the proposed change involve a 1: TSTF–2–A, Revision 1, ‘‘Relocate the 10 The proposed change revises the
significant reduction in the margin of safety? Year Sediment Cleaning of the Fuel Oil Surveillance Frequency for monitoring
Response: No. Storage Tank to Licensee Control’’ for TS [reactor coolant system] RCS temperature to
The design, operation, testing methods, pages 3.8.3–3 and 3.8.3–4 ensure the minimum temperature for
and acceptance criteria for systems, 1. Does the proposed amendment involve criticality is met. The Frequency is changed
structures, and components (SSCs), specified a significant increase in the probability or from a 30 minute Frequency when certain
in applicable codes and standards (or consequences of an accident previously conditions are met to a periodic Frequency
alternatives approved for use by the NRC) evaluated? that it is controlled in accordance with the
will continue to be met as described in the Response: No. Surveillance Frequency Control Program.
plant licensing basis (including the final The proposed change removes the The initial Frequency for this Surveillance
Surveillance Requirement for performing will be 12 hours. This will ensure that Tavg
safety analysis report and bases to TS), since
sediment cleaning of diesel fuel oil storage [average temperature] is logged at appropriate
these are not affected by changes to the tanks every 10 years from the Technical
surveillance frequencies. Similarly, there is intervals (in addition to strip chart recorders
Specifications and places it under licensee
no impact to safety analysis acceptance and computer logging of temperature). The
control. Diesel fuel oil storage tank cleaning
criteria as described in the plant licensing is not an initiator of any accident previously measurement of RCS temperature is not an
basis. To evaluate a change in the relocated evaluated. This change will have no effect on initiator of any accident previously
surveillance frequency, PPL will perform a diesel generator fuel oil quality, which is evaluated. The minimum RCS temperature
risk evaluation using the guidance contained tested in accordance with other Technical for criticality is not changed. As a result, the
in NRC approved NEI 04–10, Rev. 1 in Specifications requirements. Removing the mitigation of any accident previously
accordance with the TS SFCP. NEI 04–10, diesel fuel oil storage tank sediment cleaning evaluated is not affected.
Rev. 1, methodology provides reasonable requirements from the Technical Therefore, the proposed change does not
acceptance guidelines and methods for Specifications will have no effect on the involve a significant increase in the
ability to mitigate an accident. probability or consequences of any accident
evaluating the risk increase of proposed
Therefore, the proposed change does not previously evaluated.
changes to surveillance frequencies
involve a significant increase in the 2. Does the proposed amendment create
consistent with Regulatory Guide 1.177. probability or consequences of any accident
Therefore, the proposed changes do not the possibility of a new or different kind of
previously evaluated. accident from any accident previously
involve a significant reduction in a margin of 2. Does the proposed amendment create
safety. evaluated?
the possibility of a new or different kind of
Response: No.
accident from any accident previously
The NRC staff has reviewed the The proposed change does not involve a
evaluated?
licensee’s analysis and, based on this physical alteration to the plant (i.e., no new
Response: No.
review, it appears that the three The proposed change does not involve a or different type of equipment will be
standards of 10 CFR 50.92(c) are physical alteration to the plant (i.e., no new installed) or a change to the methods
satisfied. Therefore, the NRC staff or different type of equipment will be governing normal plant operation. The
installed) or a change to the methods changes do not alter the assumptions made
proposes to determine that the
governing normal plant operation. The in the safety analysis.
amendment request involves no
changes do not alter the assumptions made Therefore, the proposed change does not
significant hazards consideration. create the possibility of a new or different
in the safety analysis.
Attorney for licensee: Bryan A. Snapp, Therefore, the proposed change does not kind of accident from any accident
Esquire, Assoc. General Counsel, PPL create the possibility of a new or different previously evaluated.
Services Corporation, 2 North Ninth St., kind of accident from any accident 3. Does the proposed amendment involve
GENTW3, Allentown, PA 18101–1179. previously evaluated. a significant reduction in a margin of safety?
3. Does the proposed amendment involve Response: No.
NRC Branch Chief: Douglas A. a significant reduction in a margin of safety? The proposed change revises the
Broaddus. Response: No. Surveillance Frequency for monitoring RCS
The proposed change removes the temperature to ensure the minimum
Southern Nuclear Operating Company, requirement to clean sediment from the
Inc. (SNC), Docket Nos. 50–424 and 50– temperature for criticality is met. The
diesel fuel oil storage tank from the
425, Vogtle Electric Generating Plant, current, condition based Frequency
Technical Specifications and places it under
Units 1 and 2, Burke County, Georgia represents a distraction to the control room
licensee control. The margin of safety
provided by the fuel oil storage tank operator during the critical period of plant
Date of amendment request: July 18, sediment cleaning is unaffected by this startup. RCS temperature is closely
2014. A publicly-available version is in relocation because the quality of diesel fuel monitored by the operator during the
ADAMS under Accession Package No. oil is tested in accordance with other approach to criticality, and temperature is
ML14203A124. Technical Specifications requirements. recorded on charts and computer logs.
Therefore, the proposed change does not Allowing the operator to monitor
Description of amendment request: involve a significant reduction in a margin of temperature as needed by the situation and
The licensee requested 23 revisions to safety. logging RCS temperature at a periodic
the Technical Specifications (TSs). Based on the above, SNC concludes that Frequency that it is controlled in accordance
These revisions adopt various the proposed amendment does not involve a with the Surveillance Frequency Control
previously NRC-approved Technical significant hazards consideration under the Program is sufficient to ensure that the LCO
Specifications Task Force (TSTF) standards set forth in 10 CFR 50.92(c), and, [Limiting Condition for Operation] is met
Travelers. A list of the requested accordingly, a finding of ‘‘no significant while eliminating a diversion of the
revisions is included in Enclosure 1 of hazards consideration’’ is justified. operator’s attention.
the application. 2: TSTF–27–A, Revision 3, ‘‘Revise SR Therefore, the proposed change does not
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[Surveillance Requirement] Frequency for involve a significant reduction in a margin of
Basis for proposed no significant Minimum Temperature for Criticality’’ for TS safety.
hazards consideration determination: 3.4.2, TS Page 3.4.2–1 Based on the above, SNC concludes that
As required by 10 CFR 50.91(a), the 1. Does the proposed amendment involve the proposed amendment does not involve a
licensee has provided its analysis of the a significant increase in the probability or significant hazards consideration under the
issue of no significant hazards consequences of an accident previously standards set forth in 10 CFR 50.92(c), and,
consideration for each of the 24 changes evaluated? accordingly, a finding of ‘‘no significant
requested, which is presented below: Response: No. hazards consideration’’ is justified.

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11482 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

This change revises the Required Actions proposed change will not result in plant increased risk of a reactor trip. Delaying the
for LCO 3.4.5, ‘‘RCS Loops—Mode 3,’’ operation in a configuration outside the trip setpoint reduction until the power
Conditions C.2 and D.1, from ‘‘De-energize design basis. reduction has been completed and the
all control rod drive mechanisms,’’ to ‘‘Place Therefore, the proposed change does not condition is verified will minimize overall
the Rod Control System in a condition involve a significant reduction in a margin of plant risk.
incapable of rod withdrawal.’’ It also revises safety. Therefore, the proposed change does not
LCO 3.4.9, ‘‘Pressurizer,’’ Required Action Based on the above, SNC concludes that involve a significant reduction in a margin of
A.1, from requiring Reactor Trip Breakers to the proposed amendment does not involve a safety.
be open after reaching MODE 3 to ‘‘Place the significant hazards consideration under the Based on the above, SNC concludes that
Rod Control System in a condition incapable standards set forth in 10 CFR 50.92(c), and, the proposed amendment does not involve a
of rod withdrawal,’’ and to require full accordingly, a finding of ‘‘no significant significant hazards consideration under the
insertion of all rods. Inadvertent rod hazards consideration’’ is justified. standards set forth in 10 CFR 50.92(c), and,
withdrawal can be an initiator for design 7: TSTF–95–A, Revision 0, ‘‘Revise accordingly, a finding of ‘‘no significant
basis accidents or events during certain plant Completion Time for Reducing Power Range hazards consideration’’ is justified.
conditions, and therefore must be prevented High trip Setpoint from 8 to 72 Hours,’’ TS 8: TSTF–110–A, Revision 2, ‘‘Delete SR
under those conditions. The proposed 3.2.1, TS Pages 3.2.1–1 and 3.2.2–1 Frequencies Based on Inoperable Alarms,’’
Required Actions for LCO 3.4.5 and LCO TS 3.1, TS pages 3.1.4–3, 3.1.6–3, 3.2.3–1,
3.4.9 satisfy the same intent as the current 1. Does the proposed amendment involve
a significant increase in the probability or 3.2.4–4
Required Actions, which is to prevent
inadvertent rod withdrawal when an consequences of an accident previously 1. Does the proposed amendment involve
applicable Condition is not met, and is evaluated? a significant increase in the probability or
consistent with the assumptions of the Response: No. consequences of an accident previously
accident analysis. As a result, the proposed The proposed change extends the time evaluated?
allowed to reduce the Power Range Neutron Response: No.
change does not increase the probability of
Flux—High trip setpoint when Specification The proposed change removes surveillance
any accident previously evaluated. The
3.2.1, ‘‘Heat Flux Hot Channel Factor,’’ or Frequencies associated with inoperable
proposed change does not change how the
Specification 3.2.2, ‘‘Nuclear Enthalpy Rise alarms (rod position deviation monitor, rod
plant would mitigate an accident previously
Hot Channel Factor,’’ are not within their insertion limit monitor, AFD [Axial Flux
evaluated, as in both the current and
limits. Both specifications require a power Difference] monitor and QPTR [Quadrant
proposed requirements, rod withdrawal is
reduction followed by a reduction in the Power Tilt Ratio] alarm) from the Technical
prohibited.
Power Range Neutron Flux—High trip Specifications and places the actions in plant
Therefore, the proposed change does not
setpoint. Because reactor power has been administrative procedures. The subject plant
involve a significant increase in the
reduced, the reactor core power distribution alarms are not an initiator of any accident
probability or consequences of an accident previously evaluated. The subject plant
limits are within the assumptions of the
previously evaluated. alarms are not used to mitigate any accident
accident analysis. Reducing the Power Range
2. Does the proposed amendment create previously evaluated, as the control room
Neutron Flux—High trip setpoints ensures
the possibility of a new or different kind of that reactor power is not inadvertently indications of these parameters are sufficient
accident from any accident previously increased. Reducing the Power Range to alert the operator of an abnormal condition
evaluated? Neutron Flux—High trip setpoints is not an without the alarms. The alarms are not
Response: No. initiator to any accident previously credited in the accident analysis.
The proposed change provides less evaluated. The consequences of any accident Therefore, the proposed change does not
specific, but equivalent, direction on the previously evaluated with the Power Range involve a significant increase in the
manner in which inadvertent control rod Neutron Flux—High trip setpoints not probability or consequences of any accident
withdrawal is to be prevented when the reduced are no different under the proposed previously evaluated.
Conditions of LCO 3.4.5 and LCO 3.4.9 are Completion Time than under the existing 2. Does the proposed amendment create
not met. Rod withdrawal will continue to be Completion Time. Therefore, the proposed the possibility of a new or different kind of
prevented when the applicable Conditions of change does not involve a significant accident from any accident previously
LCO 3.4.5 and LCO 3.4.9 are met. There are increase in the probability or consequences evaluated?
no design changes associated with the of any accident previously evaluated. Response: No.
proposed changes, and the change does not 2. Does the proposed amendment create The proposed change does not involve a
involve a physical alteration of the plant (i.e., the possibility of a new or different kind of physical alteration to the plant (i.e., no new
no new or different type of equipment will accident from any accident previously or different type of equipment will be
be installed). The change does not alter evaluated? installed) or a change to the methods
assumptions made in the safety analysis, and Response: No. governing normal plant operation. The
is consistent with the safety analysis. The proposed change does not involve a changes do not alter the assumptions made
Therefore, the proposed change does not physical alteration to the plant (i.e., no new in the safety analysis.
create the possibility of a new or different or different type of equipment will be Therefore, the proposed change does not
kind of accident from any accident installed) or a change to the methods create the possibility of a new or different
previously evaluated. governing normal plant operation. The kind of accident from any accident
3. Does the proposed amendment involve changes do not alter the assumptions made previously evaluated.
a significant reduction in a margin of safety? in the safety analysis. 3. Does the proposed amendment involve
Response: No. Therefore, the proposed change does not a significant reduction in a margin of safety?
The proposed change provides the create the possibility of a new or different Response: No.
operational flexibility of allowing alternate, kind of accident from any accident The proposed change removes surveillance
but equivalent, methods of preventing rod previously evaluated. Frequencies associated with inoperable
withdrawal when the applicable Conditions 3. Does the proposed amendment involve alarms (rod position deviation monitor, rod
of LCO 3.4.5 and LCO 3.4.9 are met. The a significant reduction in a margin of safety? insertion limit monitor, AFD monitor and
proposed change does not affect the safety Response: No. QPTR alarm) from the Technical
analysis acceptance criteria for any analyzed The proposed change provides additional Specifications and places the actions in plant
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event, nor is there a change to any safety time before requiring the Power Range administrative procedures. The alarms are
analysis limit. The proposed change does not Neutron Flux—High trip setpoint be reduced not being removed from the plant. The
alter the manner in which safety limits, when the reactor core power distribution actions to be taken when the alarms are not
limiting safety system settings or limiting limits are not met. The manual reduction in available are proposed to be controlled under
conditions for operation are determined, nor reactor power required by the specifications licensee administrative procedures. As a
is there any adverse effect on those plant provides the necessary margin of safety for result, plant operation is unaffected by this
systems necessary to assure the this condition. Reducing the Power Range change and there is no effect on a margin of
accomplishment of protection functions. The Neutron Flux—High trip setpoints carries an safety.

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Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11483

Therefore, the proposed change does not 10: TSTF–234–A, Revision 1, ‘‘Add Action consequences of an accident previously
involve a significant reduction in a margin of for More Than One [D]RPI Inoperable,’’ TS evaluated?
safety. 3.1.7, TS Pages 3.1.7–1 and 3.1.7–2. Response: No.
Based on the above, SNC concludes that 1. Does the proposed amendment involve The proposed change revises the
the proposed amendment does not involve a a significant increase in the probability or requirements in Technical Specification
significant hazards consideration under the consequences of an accident previously 3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’
standards set forth in 10 CFR 50.92(c), and, evaluated? to clarify the operability of an AFW train
accordingly, a finding of ‘‘no significant Response: No. when it is aligned for manual steam generator
hazards consideration’’ is justified. The proposed change provides a Condition level control. The AFW System is not an
and Required Actions for more than one initiator of any design basis accident or
9: TSTF–142–A, Revision 0, ‘‘Increase the event, and therefore the proposed change
Completion Time When the Core Reactivity inoperable digital rod position indicator
(DRPI) per rod group. The DRPIs are not an does not increase the probability of any
Balance is Not Within Limit,’’ TS 3.1.2, TS accident previously evaluated. The AFW
Page 3.1.2–1 initiator of any accident previously
evaluated. The DRPIs are one indication used System is used to respond to accidents
1. Does the proposed amendment involve by operators to verify control rod insertion previously evaluated. The proposed change
a significant increase in the probability or following an accident, however other does not affect the design of the AFW
consequences of an accident previously indications are available. Therefore, allowing System, and no physical changes are made to
evaluated? a finite period to time to correct more than the plant. The proposed change does not
Response: No. one inoperable DRPI prior to requiring a significantly change how the plant would
The proposed change extends the plant shutdown will not result in a mitigate an accident previously evaluated.
Completion Time to take the Required significant increase in the consequences of Therefore, the proposed change does not
Actions when measured core reactivity is not any accident previously evaluated. involve a significant increase in the
within the specified limit of the predicted Therefore, the proposed change does not probability or consequences of an accident
values. The Completion Time to respond to involve a significant increase in the previously evaluated.
a difference between predicted and measured probability or consequences of any accident 2. Does the proposed amendment create
core reactivity is not an initiator to any previously evaluated. the possibility of a new or different kind of
accident previously evaluated. The 2. Does the proposed amendment create accident from any accident previously
consequences of an accident during the the possibility of a new or different kind of evaluated?
accident from any accident previously Response: No.
proposed Completion Time are no different
evaluated? The proposed change does not result in a
from the consequences of an accident during
Response: No. change in the manner in which the AFW
the existing Completion Time. Therefore, the
The proposed change does not involve a System provides plant protection. The AFW
proposed change does not involve a
physical alteration to the plant (i.e., no new System will continue to supply water to the
significant increase in the probability or
or different type of equipment will be steam generators to remove decay heat and
consequences of any accident previously other residual heat by delivering at least the
evaluated. installed) or a change to the methods
governing normal plant operation. The minimum required flow rate to the steam
2. Does the proposed amendment create generators. There are no design changes
the possibility of a new or different kind of changes do not alter the assumptions made
in the safety analysis. associated with the proposed changes, and
accident from any accident previously the change does not involve a physical
evaluated? Therefore, the proposed change does not
create the possibility of a new or different alteration of the plant (i.e., no new or
Response: No. different type of equipment will be installed).
The proposed change does not involve a kind of accident from any accident
previously evaluated. The change does not alter assumptions made
physical alteration to the plant (i.e., no new in the safety analysis, and is consistent with
or different type of equipment will be 3. Does the proposed amendment involve
a significant reduction in a margin of safety? the safety analysis assumptions and current
installed) or a change to the methods plant operating practice. Manual control of
governing normal plant operation. The Response: No.
The proposed change provides time to AFW level control valves is not an accident
changes do not alter the assumptions made initiator.
in the safety analysis. Therefore, the correct the condition of more than one DRPI
inoperable in a rod group. Compensatory Therefore, the proposed change does not
proposed change does not create the create the possibility of a new or different
possibility of a new or different kind of measures are required to verify that the rods
monitored by the inoperable DRPIs are not kind of accident from any accident
accident from any accident previously previously evaluated.
evaluated. moved to ensure that there is no effect on
core reactivity. Requiring a plant shutdown 3. Does the proposed amendment involve
3. Does the proposed amendment involve a significant reduction in a margin of safety?
a significant reduction in a margin of safety? with inoperable rod position indications
introduces plant risk and should not be Responses: No.
Response: No. The proposed change provides the
initiated unless the rod position indication
The proposed change provides additional operational flexibility of allowing an AFW
cannot be repaired in a reasonable period of
time to investigate and to implement train(s) to be considered operable when it is
time. As a result, the safety benefit provided
appropriate operating restrictions when not in the normal standby alignment and is
by the proposed Condition offsets the small
measured core reactivity is not within the temporarily incapable of automatic initiation,
decrease in safety resulting from continued
specified limit of the predicted values. The such as during alignment and operation for
operation with more than one inoperable
additional time will not have a significant manual steam generator level control,
DRPI.
effect on plant safety due to the provided it is capable of being manually
Therefore, the proposed change does not
conservatisms used in designing the reactor realigned to the AFW heat removal mode of
involve a significant reduction in a margin of
core and performing the safety analyses and operation. The proposed change does not
safety.
the low probability of an accident or Based on the above, SNC concludes that result in a change in the manner in which the
transient which would approach the core the proposed amendment does not involve a AFW System provides plant protection. The
design limits during the additional time. significant hazards consideration under the AFW System will continue to supply water
Therefore, the proposed change does not to the steam generators to remove decay heat
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standards set forth in 10 CFR 50.92(c), and,
involve a significant reduction in a margin of accordingly, a finding of ‘‘no significant and other residual heat by delivering at least
safety. hazards consideration’’ is justified. the minimum required flow rate to the steam
Based on the above, SNC concludes that generators. The proposed change does not
the proposed amendment does not involve a 11: TSTF–245–A, Revision 1, ‘‘AFW Train affect the safety analysis acceptance criteria
significant hazards consideration under the Operable When in Service,’’ TS 3.7.5, TS for any analyzed event, nor is there a change
standards set forth in 10 CFR 50.92(c), and, Page 3.7.5–3 to any safety analysis limit. The proposed
accordingly, a finding of ‘‘no significant 1. Does the proposed amendment involve change does not alter the manner in which
hazards consideration’’ is justified. a significant increase in the probability or safety limits, limiting safety system settings

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11484 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

or limiting conditions for operation are assumptions and current plant operating Therefore, the proposed change does not
determined, nor is there any adverse effect on practice. Operation of the PORV block valves involve a significant increase in the
those plant systems necessary to assure the is not an accident initiator. probability or consequences of any accident
accomplishment of protection functions. The Therefore, the proposed change does not previously evaluated.
proposed change will not result in plant create the possibility of a new or different 2. Does the proposed amendment create
operation in a configuration outside the kind of accident from any accident the possibility of a new or different kind of
design basis. previously evaluated. accident from any accident previously
Therefore, the proposed change does not 3. Does the proposed amendment involve evaluated?
involve a significant reduction in a margin of a significant reduction in a margin of safety? Response: No.
safety. Response: No. The proposed change does not involve a
Based on the above, SNC concludes that The proposed changes provide clarification physical alteration to the plant (i.e., no new
the proposed amendment does not involve a that separate Condition entry is allowed for or different type of equipment will be
significant hazards consideration under the each block valve. Additionally, the Actions installed) or a change to the methods
standards set forth in 10 CFR 50.92(c), and, are modified to no longer require that the governing normal plant operation. The
accordingly, a finding of ‘‘no significant PORVs be placed in manual operation when changes do not alter the assumptions made
hazards consideration’’ is justified. both block valves are inoperable and cannot in the safety analysis.
12: TSTF–247–A, Revision 0, ‘‘Provide be restored to operable status within the Therefore, the proposed change does not
Separate Condition Entry for Each [Power specified Completion Time. This preserves create the possibility of a new or different
Operated Relief Valve] PORV and Block the overpressure protection capabilities of kind of accident from any accident
Valve,’’ TS 3.4.11, TS Pages 3.4.11–1, 3.4.11– the PORVs. The proposed change does not previously evaluated.
2, 3.4.11–3 result in a change in the manner in which the 3. Does the proposed amendment involve
PORV and block valves provide plant a significant reduction in a margin of safety?
1. Does the proposed amendment involve Response: No.
a significant increase in the probability or protection. The PORVs will continue to
The proposed change modifies the
consequences of an accident previously provide overpressure protection, and the
definition of Shutdown Margin to eliminate
evaluated? block valves will continue to provide
the requirement to assume the highest worth
Response: No. isolation capability in the event a PORV is
control rod is fully withdrawn when
The proposed change revises the experiencing excessive leakage. The
calculating Shutdown Margin if it can be
requirements in Technical Specification proposed change does not affect the safety
verified by two independent means that all
3.4.11, ‘‘Pressurizer PORVs,’’ to clarify that analysis acceptance criteria for any analyzed
control rods are inserted. The additional
separate Condition entry is allowed for each event, nor is there a change to any safety
margin of safety provided by the assumption
block valve. Additionally, the Actions are analysis limit. The proposed change does not
that the highest worth control rod is fully
modified to no longer require that the PORVs alter the manner in which safety limits, withdrawn is unnecessary if it can be
be placed in manual operation when both limiting safety system settings or limiting independently verified that all controls rods
block valves are inoperable and cannot be conditions for operation are determined, nor are inserted.
restored to operable status within the is there any adverse effect on those plant Therefore, the proposed change does not
specified Completion Time. This preserves systems necessary to assure the involve a significant reduction in a margin of
the overpressure protection capabilities of accomplishment of protection functions. The safety.
the PORVs. The pressurizer block valves are proposed change will not result in plant Based on the above, SNC concludes that
used to isolate their respective PORV in the operation in a configuration outside the the proposed amendment does not involve a
event it is experiencing excessive leakage, design basis. significant hazards consideration under the
and are not an initiator of any design basis Therefore, the proposed change does not standards set forth in 10 CFR 50.92(c), and,
accident or event. Therefore the proposed involve a significant reduction in a margin of accordingly, a finding of ‘‘no significant
change does not increase the probability of safety. hazards consideration’’ is justified.
any accident previously evaluated. The Based on the above, SNC concludes that
PORV and block valves are used to respond the proposed amendment does not involve a 14: TSTF–266–A, Revision 3, ‘‘Eliminate the
to accidents previously evaluated. The significant hazards consideration under the Remote Shutdown System Table of
proposed change does not affect the design standards set forth in 10 CFR 50.92(c), and, Instrumentation and Controls,’’ TS 3.3.4, TS
of the PORV and block valves, and no accordingly, a finding of ‘‘no significant Pages 3.3.4–1, 3.3.4–3
physical changes are made to the plant. The hazards consideration’’ is justified. 1. Does the proposed amendment involve
proposed change does not change how the 13: TSTF–248–A, Revision 0, ‘‘Revise a significant increase in the probability or
plant would mitigate an accident previously Shutdown Margin Definition for Stuck Rod consequences of an accident previously
evaluated. Exception,’’ TS 1.1, TS Page 1.1–6 evaluated?
Therefore, the proposed change does not Response: No.
involve a significant increase in the 1. Does the proposed amendment involve The proposed change removes the list of
probability or consequences of an accident a significant increase in the probability or Remote Shutdown System instrumentation
previously evaluated. consequences of an accident previously and controls from the Technical
2. Does the proposed amendment create evaluated? Specifications and places them in the Bases.
the possibility of a new or different kind of Response: No. The Technical Specifications continue to
accident from any accident previously The proposed change modifies the require that the instrumentation and controls
evaluated? definition of Shutdown Margin to eliminate be operable. The location of the list of
Response: No. the requirement to assume the highest worth Remote Shutdown System instrumentation
The proposed change does not result in a control rod is fully withdrawn when and controls is not an initiator to any
change in the manner in which the PORV calculating Shutdown Margin if it can be accident previously evaluated. The proposed
and block valves provide plant protection. verified by two independent means that all change will have no effect on the mitigation
The PORVs will continue to provide control rods are inserted. The method for of any accident previously evaluated because
overpressure protection, and the block valves calculating shutdown margin is not an the instrumentation and controls continue to
will continue to provide isolation capability initiator of any accident previously be required to be operable.
in the event a PORV is experiencing evaluated. If it can be verified by two Therefore, the proposed change does not
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excessive leakage. There are no design independent means that all control rods are involve a significant increase in the
changes associated with the proposed inserted, the calculated Shutdown Margin probability or consequences of any accident
changes, and the change does not involve a without the conservatism of assuming the previously evaluated.
physical alteration of the plant (i.e., no new highest worth control rod is withdrawn is 2. Does the proposed amendment create
or different type of equipment will be accurate and consistent with the assumptions the possibility of a new or different kind of
installed). The change does not alter in the accident analysis. As a result, the accident from any accident previously
assumptions made in the safety analysis, and mitigation of any accident previously evaluated?
is consistent with the safety analysis evaluated is not affected. Response: No.

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Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11485

The proposed change does not involve a changes do not alter the assumptions made its specific TS LCO and plant configuration
physical alteration to the plant (i.e., no new in the safety analysis. will be governed by the required actions of
or different type of equipment will be Therefore, the proposed change does not those LCOs. The proposed changes are
installed) or a change to the methods create the possibility of a new or different clarifications that do not degrade the
governing normal plant operation. The kind of accident from any accident availability or capability of safety related
changes do not alter the assumptions made previously evaluated. equipment, and therefore do not create the
in the safety analysis. 3. Does the proposed amendment involve possibility of a new or different kind of
Therefore, the proposed change does not a significant reduction in a margin of safety? accident from any accident previously
create the possibility of a new or different Response: No. evaluated. There are no design changes
kind of accident from any accident The proposed change modifies the associated with the proposed changes, and
previously evaluated. Applicability of Specification 3.9.1, ‘‘Boron the changes do not involve a physical
3. Does the proposed amendment involve Concentration,’’ to clarify that the boron alteration of the plant (i.e., no new or
a significant reduction in a margin of safety? concentration limits are only applicable to different type of equipment will be installed).
Response: No. the refueling canal and the refueling cavity The changes do not alter assumptions made
The proposed change removes the list of when those volumes are attached to the RCS. in the safety analysis, and are consistent with
Remote Shutdown System instrumentation Technical Specification SR 3.0.4 requires that the safety analysis assumptions and current
and controls from the Technical Surveillances be met prior to entering the plant operating practice. Due to the
Specifications and places it in the Bases. The Applicability of a Specification. As a result, administrative nature of the changes, they
review performed by the NRC when the list the boron concentration of the refueling cannot be an accident initiator.
of Remote Shutdown System instrumentation cavity or the refueling canal must be verified Therefore, the proposed change does not
and controls is revised will no longer be to satisfy the LCO prior to connecting those create the possibility of a new or different
needed unless the criteria in 10 CFR 50.59 volumes to the RCS. The margin of safety kind of accident from any accident
are not met such that prior NRC review is provided by the refueling boron previously evaluated.
required. The Technical Specification concentration is not affected by this change 3. Does the proposed amendment involve
requirement that the Remote Shutdown as the RCS boron concentration will continue a significant reduction in a margin of safety?
System be operable, the definition of to satisfy the LCO. Response: No.
operability, the requirements of 10 CFR Therefore, the proposed change does not The proposed changes to TS 5.5.15 are
50.59, and the Technical Specifications Bases involve a significant reduction in a margin of clarifications and are editorial and
Control Program are sufficient to ensure that safety. administrative in nature. No changes are
revision of the list without prior NRC review Based on the above, SNC concludes that made the LCOs for plant equipment, the time
and approval does not introduce a significant the proposed amendment does not involve a required for the TS Required Actions to be
safety risk. significant hazards consideration under the completed, or the out of service time for the
Therefore, the proposed change does not standards set forth in 10 CFR 50.92(c), and, components involved. The proposed changes
involve a significant reduction in a margin of accordingly, a finding of ‘‘no significant do not affect the safety analysis acceptance
safety. hazards consideration’’ is justified. criteria for any analyzed event, nor is there
Based on the above, SNC concludes that 16: TSTF–273–A, Revision 2, ‘‘Safety a change to any safety analysis limit. The
the proposed amendment does not involve a Function Determination Program proposed changes do not alter the manner in
significant hazards consideration under the Clarifications,’’ TS 5.5.15, TS Page 5.5–15 which safety limits, limiting safety system
standards set forth in 10 CFR 50.92(c), and, settings or limiting conditions for operation
accordingly, a finding of ‘‘no significant 1. Does the proposed amendment involve are determined, nor is there any adverse
hazards consideration’’ is justified. a significant increase in the probability or effect on those plant systems necessary to
consequences of an accident previously assure the accomplishment of protection
15: TSTF–272–A, Revision 1, ‘‘Refueling evaluated?
Boron Concentration Clarification,’’ TS 3.9.1, functions. The proposed changes will not
Response: No. result in plant operation in a configuration
TS Page 3.9.1–1 The proposed TS changes add explanatory outside the design basis.
1. Does the proposed amendment involve text to the programmatic description of the Therefore, the proposed changes do not
a significant increase in the probability or Safety Function Determination Program involve a significant reduction in a margin of
consequences of an accident previously (SFDP) in Specification 5.5.15 to clarify in safety.
evaluated? the requirements that consideration does not Based on the above, SNC concludes that
Response: No. have to be made for a loss of power in the proposed amendment does not involve a
The proposed change modifies the determining loss of function. The Bases for significant hazards consideration under the
Applicability of Specification 3.9.1, ‘‘Boron LCO 3.0.6 is revised to provide clarification standards set forth in 10 CFR 50.92(c), and,
Concentration,’’ to clarify that the boron of the ‘‘appropriate LCO for loss of function,’’ accordingly, a finding of ‘‘no significant
concentration limits are only applicable to and that consideration does not have to be hazards consideration’’ is justified.
the refueling canal and the refueling cavity made for a loss of power in determining loss
when those volumes are attached to the of function. The changes are editorial and 17: TSTF–284–A, Revision 3, ‘‘Add ‘Met vs.
Reactor Coolant System (RCS). The boron administrative in nature, and therefore do not Perform’ to Technical Specification 1.4,
concentration of water volumes not increase the probability of any accident Frequency,’’ TS 1.4, TS 3.4, TS 3.9, TS Pages
connected to the RCS are not an initiator of previously evaluated. No physical or 1.4–1, 1.4–4, 3.4.11–3, 3.4.12–4 and 3.9.4–2
an accident previously evaluated. The ability operational changes are made to the plant. 1. Does the proposed amendment involve
to mitigate any accident previously evaluated The proposed change does not change how a significant increase in the probability or
is not affected by the boron concentration of the plant would mitigate an accident consequences of an accident previously
water volumes not connected to the RCS. previously evaluated. evaluated?
Therefore, the proposed change does not Therefore, the proposed change does not Response: No.
involve a significant increase in the involve a significant increase in the The proposed changes insert a discussion
probability or consequences of any accident probability or consequences of an accident paragraph into Specification 1.4, and several
previously evaluated. previously evaluated. new examples are added to facilitate the use
2. Does the proposed amendment create 2. Does the proposed amendment create and application of SR Notes that utilize the
the possibility of a new or different kind of the possibility of a new or different kind of terms ‘‘met’’ and ‘‘perform.’’ The changes
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accident from any accident previously accident from any accident previously also modify SRs in multiple Specifications to
evaluated? evaluated? appropriately use ‘‘met’’ and ‘‘perform’’
Response: No. Response: No. exceptions. The changes are administrative
The proposed change does not involve a The proposed changes are editorial and in nature because they provide clarification
physical alteration to the plant (i.e., no new administrative in nature and do not result in and correction of existing expectations, and
or different type of equipment will be a change in the manner in which the plant therefore the proposed change does not
installed) or a change to the methods operates. The loss of function of any specific increase the probability of any accident
governing normal plant operation. The component will continue to be addressed in previously evaluated. No physical or

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11486 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

operational changes are made to the plant. consequences of an accident previously or alter plant operating practices such that
The proposed change does not significantly evaluated? the probability of an accident previously
change how the plant would mitigate an Response: No. evaluated would be significantly increased.
accident previously evaluated. The proposed change revises Specification The proposed change does not significantly
Therefore, the proposed change does not 5.5.4, ‘‘Radioactive Effluent Controls change how the plant would mitigate an
involve a significant increase in the Program,’’ paragraph e, to describe the accident previously evaluated, and is
probability or consequences of an accident original intent of the dose projections. The bounded by the fuel handling accident (FHA)
previously evaluated. cumulative and projection of doses due to
accident analysis.
2. Does the proposed amendment create liquid releases are not an assumption in any
the possibility of a new or different kind of accident previously evaluated and have no Therefore, the proposed change does not
accident from any accident previously effect on the mitigation of any accident involve a significant increase in the
evaluated? previously evaluated. probability or consequences of an accident
Response: No. Therefore, the proposed change does not previously evaluated.
The proposed changes are administrative involve a significant increase in the 2. Does the proposed amendment create
in nature and do not result in a change in the probability or consequences of any accident the possibility of a new or different kind of
manner in which the plant operates. The previously evaluated. accident from any accident previously
proposed changes provide clarification and 2. Does the proposed amendment create evaluated?
correction of existing expectations that do the possibility of a new or different kind of Response: No.
not degrade the availability or capability of accident from any accident previously Allowing penetration flow paths to be open
safety related equipment, and therefore do evaluated? is not an initiator for any accident. The
not create the possibility of a new or different Response: No. proposed change to allow open penetration
kind of accident from any accident The proposed change does not involve a flow paths will not affect plant safety
previously evaluated. There are no design physical alteration to the plant (i.e., no new
functions or plant operating practices such
changes associated with the proposed or different type of equipment will be
that a new or different accident could be
changes, and the changes do not involve a installed) or a change to the methods
physical alteration of the plant (i.e., no new governing normal plant operation. The created. There are no design changes
or different type of equipment will be changes do not alter the assumptions made associated with the proposed changes, and
installed). The changes do not alter in the safety analysis. the change does not involve a physical
assumptions made in the safety analysis, and Therefore, the proposed change does not alteration of the plant (i.e., no new or
are consistent with the safety analysis create the possibility of a new or different different type of equipment will be installed).
assumptions and current plant operating kind of accident from any accident The change does not alter assumptions made
practice. Due to the administrative nature of previously evaluated. in the safety analysis, and is consistent with
the changes, they cannot be an accident 3. Does the proposed amendment involve the safety analysis assumptions and current
initiator. a significant reduction in a margin of safety? plant operating practice.
Therefore, the proposed change does not Response: No. Therefore, the proposed change does not
create the possibility of a new or different The proposed change revises Specification create the possibility of a new or different
kind of accident from any accident 5.5.4, ‘‘Radioactive Effluent Controls kind of accident from any accident
previously evaluated. Program,’’ paragraph e, to describe the previously evaluated.
3. Does the proposed amendment involve original intent of the dose projections. The
3. Does the proposed amendment involve
a significant reduction in a margin of safety? cumulative and projection of doses due to
a significant reduction in a margin of safety?
Response: No. liquid releases are administrative tools to
The proposed changes are administrative assure compliance with regulatory limits. Response: No.
in nature and do not result in a change in the The proposed change revises the requirement TS 3.9.4 provides measures to ensure that
manner in which the plant operates. The to clarify the intent, thereby improving the the dose consequences of a postulated FHA
proposed changes provide clarification and administrative control over this process. As inside containment are minimized. The
correction of existing expectations that do a result, any effect on the margin of safety proposed change to LCO 3.9.4 will allow
not degrade the availability or capability of should be minimal. penetration flow path(s) to be open during
safety related equipment, or alter their Therefore, the proposed change does not refueling operations under administrative
operation. The proposed changes do not involve a significant reduction in a margin of control. These administrative controls will
affect the safety analysis acceptance criteria safety. can and will be achieved in the event of an
for any analyzed event, nor is there a change Based on the above, SNC concludes that FHA inside containment, and will minimize
to any safety analysis limit. The proposed the proposed amendment does not involve a dose consequences. The proposed change is
changes do not alter the manner in which significant hazards consideration under the bounded by the existing FHA analysis. The
safety limits, limiting safety system settings standards set forth in 10 CFR 50.92(c), and, proposed change does not affect the safety
or limiting conditions for operation are accordingly, a finding of ‘‘no significant
analysis acceptance criteria for any analyzed
determined, nor is there any adverse effect on hazards consideration’’ is justified.
event, nor is there a change to any safety
those plant systems necessary to assure the 19: TSTF–312–A, Revision 1,
accomplishment of protection functions. The analysis limit. The proposed change does not
‘‘Administrative Control of Containment alter the manner in which safety limits,
proposed changes will not result in plant Penetrations,’’ TS 3.9.4, TS Page 3.9.4–1
operation in a configuration outside the limiting safety system settings or limiting
design basis. 1. Does the proposed amendment involve conditions for operation are determined, nor
Therefore, the proposed changes do not a significant increase in the probability or is there any adverse effect on those plant
involve a significant reduction in a margin of consequences of an accident previously systems necessary to assure the
safety. evaluated? accomplishment of protection functions. The
Based on the above, SNC concludes that Response: No. proposed change will not result in plant
the proposed amendment does not involve a The proposed change would allow operation in a configuration outside the
significant hazards consideration under the containment penetrations to be unisolated
design basis.
standards set forth in 10 CFR 50.92(c), and, under administrative controls during core
Therefore, the proposed change does not
alterations or movement of irradiated fuel
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accordingly, a finding of ‘‘no significant involve a significant reduction in a margin of
hazards consideration’’ is justified. assemblies within containment. The status of
containment penetration flow paths (i.e., safety.
18: TSTF–308–A, Revision 1, ‘‘Determination open or closed) is not an initiator for any Based on the above, SNC concludes that
of Cumulative and Projected Dose design basis accident or event, and therefore the proposed amendment does not involve a
Contributions in RECP [Radioactive Effluent the proposed change does not increase the significant hazards consideration under the
Controls Program],’’ TS 5.5.4, TS Page 5.5–3 probability of any accident previously standards set forth in 10 CFR 50.92(c), and,
1. Does the proposed amendment involve evaluated. The proposed change does not accordingly, a finding of ‘‘no significant
a significant increase in the probability or affect the design of the primary containment, hazards consideration’’ is justified.

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20: TSTF–314–A, Revision 0, ‘‘Require Static Response: No. The frequency of visual examinations of
and Transient FQ Measurement,’’ TS 3.1.4, The proposed change revises Specification the containment and the mode of operation
3.2.4, TS Pages 3.1.4–2, 3.2.4–1, 3.2.4–3 3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’ during which those examinations are
1. Does the proposed amendment involve to allow a 7 day Completion Time to restore performed does not affect the initiation of
a significant increase in the probability or an inoperable AFW turbine-driven pump in any accident previously evaluated. The use
consequences of an accident previously Mode 3 immediately following a refueling of NRC approved methods and frequencies
outage, if Mode 2 has not been entered. An for performing the inspections will ensure
evaluated?
inoperable AFW turbine-driven pump is not the containment continues to perform the
Response: No.
an initiator of any accident previously mitigating function assumed for accidents
The proposed change revises the Required
evaluated. The ability of the plant to mitigate previously evaluated.
Actions of Specification 3.1.4, ‘‘Rod Group
an accident is no different while in the Therefore, the proposed changes do not
Alignment Limits,’’ and Specification 3.2.4,
extended Completion Time than during the involve a significant increase in the
‘‘Quadrant Power Tilt Ratio,’’ to require
existing Completion Time. probability or consequences of an accident
measurement of both the steady state and
Therefore, the proposed change does not previously evaluated.
transient portions of the Heat Flux Hot
involve a significant increase in the 2. Does the proposed amendment create
Channel Factor, FQ(Z). This change will
probability or consequences of any accident the possibility of a new or different kind of
ensure that the hot channel factors are within
previously evaluated. accident from any accident previously
their limits when the rod alignment limits or
2. Does the proposed amendment create evaluated?
quadrant power tilt ratio are not within their
the possibility of a new or different kind of Response: No.
limits. The verification of hot channel factors
accident from any accident previously The proposed change revises the TS
is not an initiator of any accident previously
evaluated? Administrative Controls programs for
evaluated. The verification that both the
Response: No. consistency with the requirements of 10 CFR
steady state and transient portion of FQ(Z)
The proposed change does not involve a 50, paragraph 55a(g)(4) for components
are within their limits will ensure this initial
physical alteration to the plant (i.e., no new classified as Code Class CC. The proposed
assumption of the accident analysis is met
or different type of equipment will be change affects the frequency of visual
should a previously evaluated accident
installed) or a change to the methods examinations that will be performed for the
occur.
governing normal plant operation. The steel containment liner plate for the purpose
Therefore, the proposed change does not
changes do not alter the assumptions made of the Containment Leakage Rate Testing
involve a significant increase in the
in the safety analysis. Program.
probability or consequences of any accident
Therefore, the proposed change does not The proposed changes do not involve a
previously evaluated. modification to the physical configuration of
create the possibility of a new or different
2. Does the proposed amendment create the plant (i.e., no new equipment will be
kind of accident from any accident
the possibility of a new or different kind of installed) or change in the methods
previously evaluated.
accident from any accident previously governing normal plant operation. The
3. Does the proposed amendment involve
evaluated? a significant reduction in of safety? proposed changes will not impose any new
Response: No. Response: No. or different requirements or introduce a new
The proposed change does not involve a The proposed change revises Specification accident initiator, accident precursor, or
physical alteration to the plant (i.e., no new 3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’ malfunction mechanism. Additionally, there
or different type of equipment will be to allow a 7-day Completion Time to restore is no change in the types or increases in the
installed) or a change to the methods an inoperable turbine-driven AFW pump in amounts of any effluent that may be released
governing normal plant operation. The Mode 3 immediately following a refueling off-site and there is no increase in individual
changes do not alter the assumptions made outage if Mode 2 has not been entered. In or cumulative occupational exposure.
in the safety analysis. Mode 3 immediately following a refueling Therefore, the proposed changes do not
Therefore, the proposed change does not outage, core decay heat is low and the need create the possibility of a new or different
create the possibility of a new or different for AFW is also diminished. The two kind of accident from any accident
kind of accident from any accident operable motor driven AFW pumps are previously evaluated.
previously evaluated. available and there are alternate means of 3. Does the proposed amendment involve
3. Does the proposed amendment involve decay heat removal if needed. As a result, the a significant reduction in a margin of safety?
a significant reduction in a margin of safety? risk presented by the extended Completion Response: No.
Response: No. Time is minimal. The proposed changes revise the Technical
The proposed change revises the Required Therefore, the proposed change does not Specifications (TS) Administrative Controls
Actions in the Specifications for Rod Group involve a significant reduction in a margin of programs for consistency with the
Alignment Limits and Quadrant Power Tilt safety. requirements of 10 CFR 50, paragraph
Ratio to require measurement of both the Based on the above, SNC concludes that 55a(g)(4) for components classified as Code
steady state and transient portions of the the proposed amendment does not involve a Class CC. The proposed change affects the
Heat Flux Hot Channel Factor, FQ(Z). This significant hazards consideration under the frequency of visual examinations that will be
change is a correction that ensures that the standards set forth in 10 CFR 50.92(c), and, performed for the steel containment liner
plant conditions are as assumed in the accordingly, a finding of ‘‘no significant plate for the purpose of the Containment
accident analysis. hazards consideration’’ is justified. Leakage Rate Testing Program. The safety
Therefore, the proposed change does not function of the containment as a fission
22: TSTF–343–A, Revision 1, ‘‘Containment
involve a significant reduction in a margin of product barrier will be maintained.
Structural Integrity,’’ TS 5.5, TS Page 5.5–16
safety. Therefore, the proposed change does not
Based on the above, SNC concludes that 1. Does the proposed amendment involve involve a significant reduction in a margin of
the proposed amendment does not involve a a significant increase in the probability or safety.
significant hazards consideration under the consequences of an accident previously Based on the above, SNC concludes that
standards set forth in 10 CFR 50.92(c), and, evaluated? the proposed amendment does not involve a
accordingly, a finding of ‘‘no significant Response: No. significant hazards consideration under the
hazards consideration’’ is justified. The proposed change revises the Technical standards set forth in 10 CFR 50.92(c), and,
Specifications (TS) Administrative Controls
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21: TSTF–340–A, Revision 3, ‘‘Allow 7 Day accordingly, a finding of ‘‘no significant
programs for consistency with the hazards consideration’’ is justified.
Completion Time for a Turbine—Driven requirements of 10 CFR 50, paragraph
AFW Pump Inoperable,’’ TS 3.7.5, TS Page 55a(g)(4) for components classified as Code 23: TSTF–349–A, Revision 1, ‘‘Add Note to
3.7.5–1 Class CC. The proposed changes affect the LCO 3.9.5 Allowing Shutdown Cooling
1. Does the proposed amendment involve frequency of visual examinations that will be Loops Removal From Operation,’’ TS 3.9.6,
a significant increase in the probability or performed for the steel containment liner TS Page 3.9.6–1
consequences of an accident previously plate for the purpose of the Containment 1. Does the proposed amendment involve
evaluated? Leakage Rate Testing Program. a significant increase in the probability or

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11488 Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices

consequences of an accident previously standards of 10 CFR 50.92(c) are alternative manner of performing rod
evaluated? satisfied. Therefore, the NRC staff position verification. The proposed change
Response: No. proposes to determine that the does not affect the reactor protection system.
The proposed change adds an LCO Note to amendment request involves no Hence, no new failure modes are created that
LCO 3.9.6, ‘‘RHR and Coolant Circulation— would cause a new or different kind of
Low Water Level,’’ to allow securing the
significant hazards consideration. accidents from any accident previously
operating train of Residual Heat Removal Attorney for licensee: Leigh D. Perry, evaluated.
(RHR) for up to 15 minutes to support SVP & General Counsel of Operations Therefore, operation of the facility in
switching operating trains. The allowance is and Nuclear, Southern Nuclear accordance with the proposed amendments
restricted to conditions in which core outlet Operating Company, 40 Iverness Center would not create the possibility of a new or
temperature is maintained at least 10 degrees Parkway, Birmingham, AL 35201. different kind of accident from any
F below the saturation temperature, when NRC Branch Chief: Robert J. previously evaluated.
there are no draining operations, and when Pascarelli. 3. Does the proposed change involve a
operations that could reduce the reactor significant reduction in a margin of safety?
coolant system (RCS) boron concentration are Virginia Electric and Power Company, Response: No.
prohibited. Securing an RHR train to Docket Nos. 50–338 and 50–339, North The basis of TS 3.1.7 states that the
facilitate the changing of the operating train Anna Power, Unit Nos. 1 and 2, Louisa operability of the rod position indicators is
is not an initiator to any accident previously County, Virginia required to determine control rod positions
evaluated. The restrictions on the use of the and thereby ensure compliance with the
allowance ensure that an RHR train will not Date of amendment request: February control rod alignment and insertion limits.
be needed during the 15 minute period to 4, 2015. A publicly-available version is The proposed change does not alter the
mitigate any accident previously evaluated. in ADAMS under Accession No. requirement to determine rod position but
Therefore, the proposed change does not ML15041A667. provides an alternative method for
involve a significant increase in the Description of amendment request: determining the position of the affected rod.
probability or consequences of any accident The proposed license amendment As a result, the initial conditions of the
previously evaluated. requests the changes to the Technical accident analysis are preserved and the
2. Does the proposed amendment create consequences of previously analyzed
Specification (TS) TS 3.1.7, Rod
the possibility of a new or different kind of accidents are unaffected.
accident from any accident previously Position Indication, to provide an Therefore, operation of the facility in
evaluated? additional monitoring option for an accordance with the proposed amendment
Response: No. inoperable control rod position would not involve a significant reduction in
The proposed change does not involve a indicator. Specifically, the proposed a margin of safety.
physical alteration to the plant (i.e., no new changes would allow monitoring of Based on the above, Dominion concludes
or different type of equipment will be control rod drive mechanism stationary that the proposed amendment presents no
installed) or a change to the methods gripper coil voltage every eight hours as significant hazards consideration under the
governing normal plant operation. The an alternative to using the movable in standards set forth in 10 CFR 50.92(c), and,
changes do not alter the assumptions made accordingly, a finding of ‘‘no significant
core detectors every eight hours to
in the safety analysis. hazards consideration’’ is justified.
Therefore, the proposed change does not verify control rod position.
Basis for proposed no significant The NRC staff has reviewed the
create the possibility of a new or different
kind of accident from any accident hazards consideration determination: licensee’s analysis and, based on this
previously evaluated. As required by 10 CFR 50.91(a), the review, it appears that the three
3. Does the proposed amendment involve licensee has provided its analysis of the standards of 50.92(c) are satisfied.
a significant reduction in a margin of safety? issue of no significant hazards Therefore, the NRC staff proposes to
Response: No. consideration, which is presented determine that the amendment request
The proposed change adds an LCO Note to below: involves no significant hazards
LCO 3.9.6, ‘‘RHR and Coolant Circulation— consideration.
Low Water Level,’’ to allow securing the 1. Does the proposed change involve a
significant increase in the probability or
Attorney for licensee: Lillian M.
operating train of RHR to support switching Cuoco, Senior Counsel, Dominion
operating trains. The allowance is restricted consequences of an accident previously
to conditions in which core outlet evaluated? Resources Services, Inc., 120 Tredegar
temperature is maintained at least 10 degrees Response: No. Street, RS–2, Richmond, VA 23219.
F below the saturation temperature, when The proposed change provides an NRC Branch Chief: Robert Pascarelli.
there are no draining operations, and when alternative method for verifying rod position
of one rod. The proposed change meets the III. Notice of Issuance of Amendments
operations that could reduce the reactor
intent of the current specification in that it to Facility Operating Licenses and
coolant system (RCS) boron concentration are
prohibited. With these restrictions, combined ensures verification of position of the rod Combined Licenses
with the short time frame allowed to swap once every 8 hours. The proposed change During the period since publication of
operating RHR trains and the ability to start provides only an alternative method of the last biweekly notice, the
an operating RHR train if needed, the monitoring rod position and does not change
the assumptions or results of any previously
Commission has issued the following
occurrence of an event that would require amendments. The Commission has
immediate operation of an RHR train is evaluated accident.
extremely remote. Therefore, operation of the facility in determined for each of these
Therefore, the proposed change does not accordance with the proposed amendment amendments that the application
involve a significant reduction in a margin of would not involve a significant increase in complies with the standards and
safety. the probability or consequences of an requirements of the Atomic Energy Act
Based on the above, SNC concludes that accident previously evaluated. of 1954, as amended (the Act), and the
2. Does the proposed change create the
mstockstill on DSK4VPTVN1PROD with NOTICES

the proposed amendment does not involve a Commission’s rules and regulations.
significant hazards consideration under the possibility of a new or different kind of The Commission has made appropriate
standards set forth in 10 CFR 50.92(c), and, accident from any accident previously
evaluated?
findings as required by the Act and the
accordingly, a finding of ‘‘no significant Commission’s rules and regulations in
hazards consideration’’ is justified. Response: No.
The proposed change provides only an 10 CFR Chapter I, which are set forth in
The NRC staff has reviewed the alternative method of determining the the license amendment.
licensee’s analysis and, based on this position of one rod. No new accident A notice of consideration of issuance
review, it appears that the three initiators are introduced by the proposed of amendment to facility operating

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Document Created: 2015-12-18 11:55:57
Document Modified: 2015-12-18 11:55:57

Category		Regulatory Information
Collection		Federal Register
sudoc Class		AE 2.7: GS 4.107: AE 2.106:
Publisher		Office of the Federal Register, National Archives and Records Administration
Section		Notices
Action		Biweekly notice.
Dates		Comments must be filed by April 2, 2015. A request for a hearing must be filed by May 4, 2015.
Contact		Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-5411, email: [email protected]
FR Citation		80 FR 11472