[Federal Register Volume 80, Number 61 (Tuesday, March 31, 2015)]
[Notices]
[Pages 17083-17109]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-07192]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2015-0073]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
-----------------------------------------------------------------------
SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission
(NRC) is publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 5, 2015 to March 18, 2015. The last
biweekly notice was published on March 17, 2015.
DATES: Comments must be filed by April 30, 2015. A request for a
hearing must be filed by June 1, 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-2015-0073. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: 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: Kay Goldstein, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1506, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2015-0073 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-2015-0073.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2015-0073, facility name, unit
number(s), application date, and subject in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment
[[Page 17084]]
submissions to remove such information before making the comment
submissions available to the public or entering the comment submissions
into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR part 2. Interested
person(s) should consult a current copy of 10 CFR 2.309, which is
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The
NRC's regulations are accessible electronically from the NRC Library on
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment 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
[[Page 17085]]
documents over the internet, or in some cases to mail copies on
electronic storage media. Participants may not submit paper copies of
their filings unless they seek an exemption in accordance with the
procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten 10 days prior to the filing deadline, the participant should
contact the Office of the Secretary by email at [email protected],
or by telephone at 301-415-1677, to request (1) a digital
identification (ID) certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's public
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)-(iii).
For further details with respect to 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.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 24, 2014. A publicly-
[[Page 17086]]
available version is in ADAMS under Accession No. ML14330A327.
Description of amendment request: The proposed amendments would
modify the Technical Specifications (TS) to correct non-conservative
setpoints. Specifically, modify the Allowable Value parameter and the
Nominal Trip Setpoint for the TS 3.3.2 Table 3.3.2-1, ``Engineered
Safety Feature Actuation System Instrumentation'' function for
Auxiliary Feedwater Loss of Offsite Power (Function 6.d.) and for the
TS 3.3.5 Loss of Voltage function in Surveillance Requirement (SR)
3.3.5.2. As part of the change, the licensee is also proposing to add
the applicable footnotes in accordance with TSTF-493, Revision 4,
``Clarify Application of Setpoint Methodology for LSSS [limiting safety
system set point] Functions.''
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 and staff's changes/additions
are provided in [ ]:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Duke Energy requests NRC review and approval to revise the
Allowable Value parameter for the Technical Specification (TS) 3.3.2
Table 3.3.2-1, ``Engineered Safety Feature Actuation System
Instrumentation'' function for Auxiliary Feedwater Loss of Offsite
Power (Function 6.d.) and for the TS 3.3.5 Loss of Voltage function
in Surveillance Requirement (SR) 3.3.5.2 in order to make this
parameter more restrictive. The existing parameter was determined to
be non-conservative and this parameter is presently classified as
Operable But Degraded in the Catawba Corrective Action Program. In
addition, the Nominal Trip Setpoint parameter for this function is
being slightly lowered in order to gain additional margin. Finally,
as part of this License Amendment Request (LAR), applicable
footnotes are also being added to the affected TS 3.3.2 function in
accordance with TS Task Force Traveler [(TSTF)] TSTF-493, Revision
4, ``Clarify Application of Setpoint Methodology for LSSS
Functions.'' The more restrictive Allowable Value will preclude the
potential for a double sequencing event to occur under the condition
of a Loss of Coolant Accident (LOCA) load sequencer actuation with a
pre-existing degraded voltage condition on the essential buses.
These proposed changes will not increase the probability of
occurrence of any design basis accident since the affected function,
in and of itself, cannot initiate an accident. Should a LOCA occur,
the proposed changes will ensure that the sequencer operates
properly in order to mitigate the consequences of the event.
Appropriate calculations were developed to substantiate the revised
TS parameters proposed in this LAR. There will be no impact on the
source term or pathways assumed in accidents previously evaluated.
No analysis assumptions will be violated and there will be no
adverse effects on onsite or offsite doses as the result of an
accident. Adoption of the TSTF-493 footnotes for the respective SRs
will ensure that the function's channels will continue to behave in
accordance with safety analysis assumptions and the channel
performance assumptions in the setpoint methodology.
Therefore, the proposed amendments 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 amendments do not change the methods governing
normal plant operation; nor are the methods utilized to respond to
plant transients altered. In addition, the proposed changes to the
affected TS parameters and the adoption of the TSTF-493 footnotes
will not create the potential for any new initiating events or
transients to occur in the actual physical plant.
Therefore, the proposed amendments 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 the margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
The proposed changes will assure the acceptable operation of the
affected function under all postulated transient and accident
conditions. This will ensure that all applicable design and safety
limits are satisfied such that the fission product barriers will
continue to perform their design functions.
Therefore, the proposed amendments do not involve a significant
reduction in a margin of safety.
Based on the preceding discussion, Duke Energy concludes that
the proposed amendments do not involve a significant hazards
consideration under the standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ``no significant hazards consideration''
is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Robert J. Pascarelli.
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 amendment request: March 14, 2014. A publicly-available
version is in ADAMS under Accession No. ML14078A037.
Description of amendment request: The amendment would revise the
Technical Specifications (TS) for the Inservice Testing Program to
reflect the current edition of the American Society of Mechanical
Engineers (ASME) Code that is referenced in 10 CFR 50.55a(b).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change corrects a typographical error in TS 5.5.8,
``Reactor Coolant Pump Flywheel Inspection Program,'' and revises TS
5.5.9, ``lnservice Testing Program,'' for consistency with the
requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing
of pumps and valves which are classified as ASME Code Class 1, Class
2 and Class 3. The proposed change incorporates revisions to the
ASME Code that result in a net improvement in the measures for
testing pumps and valves.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. The proposed change does not involve the addition or removal
of any equipment, or any design changes to the facility.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change corrects a typographical error in TS 5.5.8,
``Reactor Coolant Pump Flywheel Inspection Program,'' and revises TS
5.5.9, ``lnservice Testing Program,'' for consistency with the
requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing
of pumps and valves which are classified as ASME Code Class 1, Class
2 and Class 3. The proposed change incorporates revisions to the
ASME Code that result in a net improvement in the measures for
testing pumps and valves.
The proposed change does not involve a modification to the
physical configuration of
[[Page 17087]]
the plant (i.e., no new equipment will be installed), nor does it
involve a change in the methods governing normal plant operation.
The proposed change 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 offsite and there is no increase in individual or
cumulative occupational exposure.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change corrects a typographical error in TS 5.5.8,
``Reactor Coolant Pump Flywheel Inspection Program,'' and revises TS
5.5.9, ``lnservice Testing Program,'' for consistency with the
requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing
of pumps and valves which are classified as ASME Code Class 1, Class
2 and Class 3. The proposed change incorporates revisions to the
ASME Code that result in a net improvement in the measures for
testing pumps and valves. The safety function of the affected pumps
and valves will be maintained. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Date of amendment request: November 21, 2014. A publicly-available
version is in ADAMS under Accession No. ML14325A520.
Description of amendment request: The amendment would change the
GGNS Technical Specification (TS) 2.1.1, ``Reactor Core SLs [Safety
Limits].'' Specifically, the change would revise the Minimum Critical
Power Ratio (MCPR) SL stated in TS 2.1.1.2 for two-loop operation from
greater than or equal to (>=) 1.11 to >= 1.15. Additionally, the change
would revise the MCPR SL stated in TS 2.1.1.2 for single-loop operation
from >= 1.14 to >= 1.15.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Bases to TS 2.1.1.2 states that: ``The MCPR SL ensures
sufficient conservatism in the operating MCPR limit that, in the
event of an AOO [Anticipated Operational Occurrence] from the
limiting condition of operation, at least 99.9% of the fuel rods in
the core would be expected to avoid boiling transition.
This condition is met in that the GGNS Cycle 20 (C20) MCPR SL
evaluation was performed in accordance with Reference 4 [NEDE-24011-
P-A, ``General Electric Standard Application for Reactor Fuel
(GESTAR-II'')]. The resulting values continue to ensure the
conservatism described in the Bases to TS 2.1.1.2. The proposed
changes also continue to ensure sufficient conservatism in the
operating MCPR limit. The MCPR operating limits are presented and
controlled in accordance with the GGNS Core Operating Limits Report
(COLR).
The requested Technical Specification change does not involve
any plant modifications or operational changes that could affect
system reliability or performance or that could affect the
probability of operator error. The requested change does not affect
any postulated accident precursors, any accident mitigating systems,
or introduce any new accident initiation mechanisms.
Therefore, the proposed change to increase the MCPR SL values
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any new modes of operation,
any changes to setpoints, or any plant modifications. The proposed
change to the MCPR SL accounts for requirements specified in the NRC
Safety Evaluation limitations and conditions associated with NEDC-
33173P [``Applicability of GE Methods to Expanded Operating
Domains''] and NEDC-33006P [``Licensing Topical Report--General
Electric Boiling Water Reactor Maximum Extended Load Line Limit
Analysis Plus'']. Compliance with the criterion for incipient
boiling transition continues to be ensured. The core operating
limits will continue to be developed using NRC approved methods. The
proposed [MCPR SL] does not result in the creation of any new
precursors to an accident.
Therefore, the proposed change does not create of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The MCPR SLs have been evaluated in accordance with Global
Nuclear Fuels NRC-approved cycle-specific safety limit methodology
to ensure that during normal operation and during AOO's, at least
99.9% of the fuel rods in the core are not expected to experience
transition boiling. The proposed change to the [MCPR SL] accounts
for requirements specified in the NRC Safety Evaluation limitations
and conditions associated with NEDC-33173P and NEDC-33006P, which
result in additional margin above that specified in the TS Bases.
Therefore, the proposed change to the MCPR SL does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Meena K. Khanna.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Date of amendment request: November 21, 2014, as supplemented by
letter dated February 18, 2015. Publicly-available versions are in
ADAMS under Accession Nos. ML14325A752 and ML15049A536, respectively.
Description of amendment request: The proposed amendment would
revise GGNS's license basis to adopt a single fluence methodology.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change adopts a single flux methodology. While
Chapter 15, Accident Analysis, of the Standard Review Plan (NUREG-
0800, Standard Review Plan for the Review of Safety Analysis Reports
for Nuclear Power Plants) assumes the pressure vessel does not fail,
the flux methodology is not an initiator to any accident previously
evaluated. Accordingly, the proposed change
[[Page 17088]]
to the adoption of the flux methodology has no effect on the
probability of any 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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change adopts a flux methodology. The change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. The change does not alter
assumptions made in the safety analysis regarding fluence.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adopts a single fluence methodology. The
proposed change does not alter the manner in which safety limits,
limiting safety system settings or limiting conditions for operation
are determined. The proposed change ensures that the methodology
used for fluence is in compliance with RG 1.190 requirements.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Meena K. Khanna.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Date of amendment request: August 19, 2014. A publicly-available
version is in ADAMS under Accession No. ML14231A902.
Description of amendment request: The proposed amendment would
increase the technical specification (TS) surveillance requirement (SR)
3.7.9.2 allowable temperature to less than or equal to
102[emsp14][deg]F [degree Fahrenheit].
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the Proposed Change Involve a Significant Increase in
the Probability or Consequences of an Accident Previously Evaluated?
Response: No.
The likelihood of a malfunction of any systems, structures or
components (SSCs) supported by the UHS [ultimate heat sink] is not
significantly increased by increasing the allowable Ultimate Heat
Sink (UHS) temperature from <=100[emsp14][deg]F to
<=102[emsp14][deg]F. The UHS provides a heat sink for process and
operating heat from safety related components during a transient or
accident, as well as during normal operation. The proposed change
does not make any physical changes to any plant SSCs, nor does it
alter any of the assumptions or conditions upon which the UHS is
designed. The UHS is not an initiator of any analyzed accident. All
equipment supported by the UHS has been evaluated to demonstrate
that their performance and operation remains as described in the
UFSAR [updated final safety analysis report] with no increase in
probability of failure or malfunction.
The SSCs credited to mitigate the consequences of postulated
design basis accidents remain capable of performing their design
basis function. The change in maximum UHS temperature has been
evaluated using the UFSAR described methods to demonstrate that the
UHS remains capable of removing normal operating and post-accident
heat. The change in UHS temperature and resulting containment
response following a postulated design basis accident has been
demonstrated to not be impacted. Additionally, all the UHS supported
equipment, credited in the accident analysis to mitigate an
accident, has been shown to continue to perform their design
function as described in the UFSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the Proposed Change Create the Possibility of a New or
Different Kind of Accident from any Accident Previously Evaluated?
Response: No.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not introduce any new modes of plant
operation, change the design function of any SSC, change the mode of
operation of any SSC, or change any actions required when the TS
limit is exceeded. There are no new equipment failure modes or
malfunctions created as affected SSCs continue to operate in the
same manner as previously evaluated and have been evaluated to
perform as designed at the increased UHS temperature and as assumed
in the accident analysis. Additionally, accident initiators remain
as described in the UFSAR and no new accident initiators are
postulated as a result of the increase in UHS temperature.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the Proposed Change Involve a Significant Reduction in a
Margin of Safety?
Response: No.
The proposed change continues to ensure that the maximum
temperature of the cooling water supplied to the plant SSCs during a
UHS design basis event remains within the evaluated equipment limits
and capabilities assumed in the accident analysis. The proposed
change does not result in any changes to plant equipment function,
including setpoints and actuations. All equipment will function as
designed in the plant safety analysis without any physical
modifications. The proposed change does not alter a limiting
condition for operation, limiting safety system setting, or safety
limit specified in the Technical Specifications.
The proposed change does not adversely impact the UHS inventory
required to be available for the UFSAR described design basis
accident involving the worst case 30-day period including losses for
evaporation and seepage to support safe shutdown and cooldown of
both Braidwood Station units. Additionally, the structural integrity
of the UHS is not impacted and remains acceptable following the
change, thereby ensuring that the assumptions for both UHS
temperature and inventory remain valid.
Therefore, since there is no adverse impact of this change on
the Braidwood Station safety analysis, there is no reduction in the
margin of safety of the plant.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate
Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-454 and STN
50-455, Byron Station, Units 1 and 2, Ogle County, Illinois
Date of amendment request: November 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14328A800.
Description of amendment request: The proposed amendment would
revise Condition I and surveillance requirement (SR) 3.7.9.3 associated
with technical specification (TS) Section 3.7.9, ``Ultimate Heat Sink
(UHS),'' to reflect the current design basis flood level.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 17089]]
consideration, which is presented below:
EGC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92(c), ``Issuance of
amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to revise TS 3.7.9, Condition I and SR
3.7.9.3 will ensure the operability of the SX [service water] makeup
pumps to meet TS 3.7.9 LCO [Limiting Condition for Operation]
requirement. The proposed change does not result in any physical
changes to safety related structures, systems, or components. The
probability of a flood at the river screen house (RSH) is unchanged.
Since the UHS itself is not an accident initiator, the proposed
change does not impact the initiators or assumptions of analyzed
accidents, nor do they impact the mitigation of accidents or
transient events. Consequently, the proposed change does not
increase the probability of occurrence for any accident previously
evaluated.
The proposed change will ensure that actions to verify
operability of the deep well pumps will be taken prior to the
potential for the SX makeup pumps to be adversely affected by the
combined event flood high river level. Therefore, the UHS will be
capable of performing its functions to mitigate accidents by serving
as the heat sink for safety related equipment. Thus, the proposed
change does not increase the consequences of any 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 to revise TS 3.7.9, Condition I and SR
3.7.9.3 does not change the design function or operation of the SX
makeup pumps. The proposed change does not change or introduce the
possibility of any new or different type of equipment, modes of
system operation, failure mechanisms, malfunctions, or accident
initiators. The proposed change to lower the river level value at
which action is taken to verify basin levels and deep well pumps are
ready to perform the UHS makeup function in the place of the SX
makeup pumps will not affect the operation or function of the UHS or
the deep well pumps.
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 to revise TS 3.7.9, Condition I and SR
3.7.9.3 reestablishes the margin between the design bases combined
event flood level and TS 3.7.9, Condition I action level for high
river level. The proposed change will ensure the operability of the
SX makeup pumps to meet TS 3.7.9 LCO and do not affect the ability
of the SX makeup pumps to provide the safety related source makeup
to the UHS.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, EGC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and accordingly, a finding
of no significant hazards consideration is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: December 22, 2014. A publicly-available
version is in ADAMS under Accession No. ML14357A085.
Description of amendment request: The proposed amendment modifies
the technical specifications (TSs) to add a new Limiting Condition for
Operation (LCO) 3.10.8 to specifically permit inservice leakage and
hydrostatic testing at reactor coolant system (RCS) temperatures
greater than the average reactor coolant temperature for MODE 4 with
the reactor shutdown. In addition, the proposed amendment includes an
expanded scope of LCO 3.10.8 consistent with the NRC-approved Revision
0 of Technical Specification Task Force (TSTF) Improved Standard
Technical Specification Change Traveler, TSTF-484, ``Use of TS 3.10.1
for Scram Time Testing Activities'' available in ADAMS under Accession
No. ML062990425.
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:
EGC [Exelon Generation Company] has evaluated the proposed
changes, using the criteria in 10 CFR 50.92, and has determined that
the proposed changes do not involve a significant hazards
consideration. The following information is provided to support a
finding of no significant hazards consideration.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes will not result in a significant change in
the stored energy in the reactor vessel during the performance of
the testing. The probability of an accident is not significantly
increased because the proposed changes will not alter the method by
which inservice leakage and hydrostatic testing is performed or
significantly change the temperatures and pressures achieved to
perform the test.
The consequences of previously evaluated accidents are not
significantly increased because the required testing conditions
provide adequate assurance that the consequences of a steam leak
will be conservatively bounded by the consequences of the postulated
main system line break outside of primary containment. Under these
proposed changes, the secondary containment, standby gas treatment
system, and associated initiation instrumentation are required to be
operable during the performance of inservice leakage and hydrostatic
testing and would be capable of mitigating any airborne
radioactivity or steam leaks that could occur. In addition, the
required Emergency Core Cooling subsystems will be more than
adequate to ensure that a significant increase in consequences will
not occur by ensuring that the potential for failed fuel and a
subsequent increase in coolant activity above Technical
Specification limits are minimized.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
As the accumulated neutron fluence on the reactor vessel
increases, the Pressure-Temperature Limits in TS 3.4.9 for DNPS
[Dresden Nuclear Power Station] and QCNPS [Quad Cities Nuclear Power
Station and TS [technical specification] 3.4.11 for LSCS [LaSalle
County Station] may eventually require that inservice leakage and
hydrostatic testing be conducted at RCS [reactor coolant system]
temperatures greater than the average reactor coolant temperature
for MODE 4 with the reactor shutdown. However, even with the
required minimum reactor coolant temperatures less than or equal to
the average reactor coolant temperature for MODE 4 with the reactor
shutdown, maintaining RCS
[[Page 17090]]
temperatures within a small band during testing can be impractical.
The proposed changes will not result in a significant change in the
stored energy in the reactor vessel during the performance of the
testing nor will it alter the way inservice leakage and hydrostatic
testing is performed or significantly change the temperatures and
pressures achieved to perform the testing.
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 a
margin of safety?
Response: No.
The proposed changes and additions result in increased system
operability requirements above those that currently exist during the
performance of inservice leakage and hydrostatic testing. The
incremental increase in stored energy in the vessel during testing
will be conservatively bounded by the consequences of the postulated
main steam line break outside of primary containment and analyzed
margins of safety are unchanged.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
EGC has reviewed the no significant hazards determination
published on August 21, 2006 (71 FR 48561) [for Technical
Specification Task Force traveler TSTF-484]. The no significant
hazards determination was made available on October 27, 2006 (71 FR
63050) as part of the CLIIP [Consolidated Line Item Improvement
Process] Notice of Availability. EGC has concluded that the
determination presented in the notice is applicable to DNPS, Units 2
and 3; LSCS, Units 1 and 2; and QCNPS, Units 1 and 2; and the
determination is hereby incorporated by reference to satisfy the
requirements of 10 CFR 50.91(a).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Bradley Fewell, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket No. 50-373 and 50-374, LaSalle
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois
Date of Amendment Request: January 12, 2015. A publicly-available
version is in ADAMS under Accession No. ML15012A544.
Description of amendment request: The proposed amendment would
delete the limiting condition for operation (LCO) Note for Technical
Specification (TS) Section 3.5.1, ``ECCS [emergency core cooling
system]--Operating.'' The current Note allows the licensee to consider
the low pressure coolant injection (LPCI) subsystem associated with the
residual heat removal (RHR) system to be OPERABLE under specified
conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
No physical changes to the facility will occur as a result of
this proposed amendment. The proposed change will not alter the
physical design. Current TS note could make LSCS susceptible to
potential water hammer in the RHR system if in the SDC [shutdown
cooling] Mode of RHR in Mode 3 when swapping from the SDC to LPCI
mode of RHR. The proposed LAR [license amendment request] will
eliminate the risk for cavitation of the pump and voiding in the
suction piping, thereby avoiding potential to damage the RHR system,
including water hammer.
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 alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Accordingly, the change does not introduce any new
accident initiators, nor does it reduce or adversely affect the
capabilities of any plant structure, system, or component to perform
their safety function. Deletion of the TS note is appropriate
because current TSs could put the plant at risk for potential
cavitation of the pump and voiding in the suction piping, resulting
in potential to damage the RHR system, including water hammer.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change conforms to NRC regulatory guidance
regarding the content of plant Technical Specifications. The
proposed change does not alter the physical design, safety limits,
or safety analysis assumptions associated with the operation of the
plant.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above evaluation, EGC [Exelon Generation Company,
LLC] concludes that the proposed amendment does not involve a
significant hazards consideration under the standards set forth in
10 CFR 50.92(c), and, according a finding of no significant hazards
consideration is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Exelon Generation Company,
LLC, 4300 Winfield Road, Warrenville, IL, 60555.
Branch Chief: Travis L. Tate.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: December 31, 2014. A publicly-available
version is in ADAMS under Accession No. ML14365A080.
Description of amendment request: The proposed amendment would
revise the frequency for the technical specification surveillance to
verify that each containment spray system nozzle is unobstructed from a
frequency of 10 years to an event-based frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The containment spray system and its spray nozzles are not
accident initiators and therefore the proposed change does not
involve a significant increase in the probability of an accident.
The revised surveillance requirement will require event-based
frequency verification in lieu of a fixed frequency verification.
The proposed change does not have a detrimental impact on the
integrity of any plant structure, system, or component that may
initiate an analyzed event. The proposed change will not alter the
operation or otherwise increase the failure probability of any plant
equipment that can initiate an analyzed accident. Because the system
will continue to be available to perform its accident mitigation
function, the consequences of accidents 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?
[[Page 17091]]
Response: No.
The proposed change will not physically alter the plant (no new
or different type of equipment will be installed) or change the
methods governing normal plant operation. The proposed change does
not introduce new accident initiators or impact assumptions made in
the safety analysis. Testing requirements continue to demonstrate
that the limiting conditions for operation are met and the system
components are functional.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The safety function of the CSS [containment spray system] is to
spray water into the containment atmosphere in the event of a loss-
of-coolant accident to prevent containment pressure from exceeding
the design value and to remove fission products from the containment
atmosphere.
The CSS is not susceptible to corrosion-induced obstruction or
obstruction from sources external to the system. Maintenance
activities that unexpectedly introduce unretrievable foreign
material into the system would require subsequent verification to
ensure there is no nozzle blockage. The spray header nozzles are
expected to remain unblocked and available in the event that a
safety function is required. Therefore, the capacity of the system
would remain unaffected. The proposed change does not relax any
criteria used to establish safety limits and will not relax any
safety system settings. The safety analysis acceptance criteria are
not affected by this change.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Travis L. Tate.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: December 19, 2014. A publicly-available
version is in ADAMS under Accession No. ML14353A349.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to adopt performance-based
Type C testing for the reactor containment, which would allow for
extended test intervals for Type C valves up to 75 months, and corrects
an editorial issue in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment adopts the NRG-accepted guidelines of
[Nuclear Energy Institute] NEI 94-01, Revision 3-A, ``Industry
Guideline for Implementing Performance-Based Option of 10 CFR part
50, Appendix J,'' for [Davis-Besse Nuclear Power Station] DBNPS
performance-based Type C containment isolation valve testing.
Revision 3-A of NEI 94-01 allows, based on previous valve leak test
performance, an extension of Type C containment isolation valve leak
test intervals. Since the change involves only performance-based
Type C testing, the proposed amendment does not involve either a
physical change to the plant or a change in the manner in which the
plant is operated or controlled.
Implementation of these guidelines continues to provide adequate
assurance that during design basis accidents, the components of the
primary containment system will limit leakage rates to less than the
values assumed in the plant safety analyses.
The proposed amendment will not change the leakage rate
acceptance requirements. As such, the containment will continue to
perform its design function as a barrier to fission product
releases.
Therefore, the proposed amendment does not significantly
increase 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 amendment to revise the extended frequency
performance-based Type C testing program does not change the design
or operation of structures, systems, or components of the plant.
The proposed amendment would continue to ensure containment
operability and would ensure operation within the bounds of existing
accident analyses. There are no accident initiators created or
affected by the proposed amendment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment to revise the extended frequency
performance-based Type C testing program does not affect plant
operations, design functions, or any analysis that verifies the
capability of a structure, system, or component of the plant to
perform a design function. In addition, this change does not affect
safety limits, limiting safety system setpoints, or limiting
conditions for operation. The specific requirements and conditions
of the Technical Specification Containment Leakage Rate Testing
Program exist to ensure that the degree of containment structural
integrity and leak-tightness that is considered in the plant safety
analysis is maintained.
The overall containment leak rate limit specified by Technical
Specifications is maintained, thus ensuring the margin of safety in
the plant safety analysis is maintained. The design, operation,
testing methods, and acceptance criteria for Type A, Type B, and
Type C containment leakage tests specified in applicable codes and
standards would continue to be met with the acceptance of this
proposed change, since these are not affected by this revision to
the performance-based containment testing program.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Travis L. Tate.
Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: November 14, 2014, as supplemented by a
letter dated February 12, 2015. Publicly-available versions are in
ADAMS under Accession Nos. ML14324A209, and ML15050A247, respectively.)
Description of amendment request: The proposed amendments would
replace the current Donald C. Cook Nuclear Plant (CNP) Units 1 and 2
technical specifications (TSs) limit on reactor coolant system (RCS)
gross specific activity with a new limit on RCS noble gas specific
activity. The noble gas specific activity limit would be based on a new
DOSE EQUIVALENT XE-133 definition that would replace the current E-Bar
average disintegration energy definition. In addition, the current DOSE
EQUIVALENT I-131 definition would be revised to allow the use of
additional thyroid dose conversion factors. The proposed RCS specific
activity changes are consistent with NRC-approved Industry Technical
[[Page 17092]]
Specification Task Force (TSTF) Standard Technical Specification change
traveler, TSTF-490, Revision 0, ``Deletion of E-Bar Definition and
Revision to Reactor Coolant System Specific Activity Technical
Specification,'' with deviations. Additionally, the proposed amendments
would revise the CNP Units 1 and 2 licensing basis and TSs to adopt the
alternative source term (AST) as allowed in 10 CFR 50.67. The proposed
amendments represent full implementation of the AST as described in the
NRC's Regulatory Guide 1.183, ``Alternative Radiological Source Terms
for Evaluating Design Basis Accidents at Nuclear Power Reactors,''
Revision 0.
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. The licensee concluded that the no significant hazards
consideration determination published on March 19, 2007 (72 FR 12838),
``Notice of Availability of the Model Safety Evaluation,'' is
applicable. This determination is presented below, along with the
licensee's analysis of the implementation of the AST.
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
Reactor coolant specific activity is not an initiator for any
accident previously evaluated. The Completion Time when primary
coolant gross activity is not within limit is not an initiator for
any accident previously evaluated. The current variable limit on
primary coolant iodine concentration is not an initiator to any
accident previously evaluated. As a result, the proposed change does
not significantly increase the probability of an accident. The
proposed change will limit primary coolant noble gases to
concentrations consistent with the accident analyses. The proposed
change to the Completion Time has no impact on the consequences of
any design basis accident since the consequences of an accident
during the extended Completion Time are the same as the consequences
of an accident during the Completion Time. As a result, the
consequences of any accident previously evaluated are not
significantly increased.
There are no physical changes to the plant being introduced by
the proposed changes to the accident source term. Implementation of
AST and the associated proposed TS changes and new atmospheric
dispersion factors have no impact on the probability for initiation
of any DBAs [Design Basis Accidents]. Once the occurrence of an
accident has been postulated, the new accident source term and
atmospheric dispersion factors are an input to analyses that
evaluate the radiological consequences. The proposed changes do not
involve a revision to the design or manner in which the facility is
operated that could increase the probability of an accident
previously evaluated in Chapter 14 of the UFSAR.
Based on the AST analyses, there are no proposed changes to
performance requirements and no proposed revision to the parameters
or conditions that could contribute to the initiation of an accident
previously discussed in Chapter 14 of the UFSAR. Plant-specific
radiological analyses have been performed using the AST methodology
and new X/Qs have been established. Based on the results of these
analyses, it has been demonstrated that the CR [control room] and
off-site dose consequences of the limiting events considered in the
analyses meet the regulatory guidance provided for use with the AST,
and the doses are within the limits established by 10 CFR 50.67.
Therefore, it is concluded that the proposed amendment does not
involve a significant increase in the probability or the
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 in specific activity limits does not alter
any physical part of the plant nor does it affect any plant
operating parameter. The change does not create the potential for a
new or different kind of accident from any previously calculated.
No new modes of operation are introduced by the proposed
changes. The proposed changes will not create any failure mode not
bounded by previously evaluated accidents. Implementation of AST and
the associated proposed TS changes and new X/Qs have no impact to
the initiation of any DBAs. These changes do not affect the design
function or modes of operation of structures, systems and components
in the facility prior to a postulated accident. Since structures,
systems and components are operated no differently after the AST
implementation, no new failure modes are created by this proposed
change. The alternative source term change itself does not have the
capability to initiate accidents.
Consequently, the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change revises the limits on noble gas
radioactivity in the primary coolant. The proposed change is
consistent with the assumptions in the safety analyses and will
ensure the monitored values protect the initial assumptions in the
safety analyses.
The AST analyses have been performed using approved
methodologies to ensure that analyzed events are bounding and safety
margin has not been reduced. Also, new X/Qs, which are based on site
specific meteorological data, were calculated in accordance with the
guidance of RG 1.194 to utilize more recent data and improved
calculational methodologies. The dose consequences of these limiting
events are within the acceptance criteria presented in 10 CFR 50.67.
Thus, by meeting the applicable regulatory limits for AST, there is
no significant reduction in a margin of safety. Therefore, because
the proposed changes continue to result in dose consequences within
the applicable regulatory limits, the proposed amendment does not
involve a significant reduction in 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 amendments
requested involve no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David L. Pelton.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2, Somervell
County, Texas
Date of amendment request: January 28, 2015. A publicly-available
version is in ADAMS under Accession No. ML15036A032.
Description of amendment request: The amendment would revise
Technical Specification (TS) 5.5.16, ``Containment Leakage Rate Testing
Program,'' for CPNPP, Units 1 and 2, to allow an increase in the 10 CFR
part 50, appendix J, ``Primary Reactor Containment Leakage Testing for
Water-Cooled Power Reactors,'' Type A Integrated Leak Rate Test (ILRT)
interval from a 10-year frequency to a maximum of 15 years and the
extension of the containment isolation valves leakage Type C tests from
its current 60-month frequency to 75 months in accordance with Nuclear
Energy Institute (NEI) 94-01, Revision 3-A, ``Industry Guidance for
Implementing Performance Based Option of 10 CFR part 50, appendix J,''
July 2012 (ADAMS Accession No. ML12221A202), and conditions and
limitations specified in NEI 94-01, Revision 2-A, ``Industry Guidance
for Implementing Performance Based Option of 10 CFR part 50, appendix
J,'' October 2008 (ADAMS Accession No. ML100620847), in addition to
limitations and conditions of NEI 94-01, Revision 3-A. The proposed
change would also delete the listing of one-time exceptions previously
granted to ILRT frequencies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 17093]]
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the TS involves the extension of the
CPNPP, Units 1 and 2 Type A containment test interval to 15 years
and the extension of the Type C test interval to 75 months. The
current Type A test interval of 120 months (10 years) would be
extended on a permanent basis to no longer than 15 years from the
last Type A test. The current Type C test interval of 60 months for
selected components would be extended on a performance basis to no
longer than 75 months. Extensions of up to nine months (total
maximum interval of 84 months for Type C tests) are permissible only
for non-routine emergent conditions. The proposed extension does not
involve either a physical change to the plant or a change in the
manner in which the plant is operated or controlled. The containment
is designed to provide an essentially leak tight barrier against the
uncontrolled release of radioactivity to the environment for
postulated accidents. The containment and the testing requirements
invoked to periodically demonstrate the integrity of the containment
exist to ensure the plant's ability to mitigate the consequences of
an accident, and do not involve the prevention or identification of
any precursors of an accident. The change in dose risk for changing
the Type A test frequency from three-per-ten years to once-per-
fifteen-years, measured as an increase to the total integrated dose
risk for all internal events accident sequences for CPNPP, of 1.00E-
02 person rem/yr [roentgen equivalent man per year] to 6.51 person-
rem/yr for Unit 1 and 6.53 person-rem/yr for Unit 2 using the EPRI
[Energy Power Research Institute] guidance with the base case
corrosion included. Therefore, this proposed extension does not
involve a significant increase in the probability of an accident
previously evaluated.
As documented in NUREG-1493 [, ``Performance-Based Containment
Leak-Test Program: Draft Report for Comment,'' January 1995 (not
publicly available)], Type B and C tests have identified a very
large percentage of containment leakage paths, and the percentage of
containment leakage paths that are detected only by Type A testing
is very small. The CPNPP, Units 1 and 2 Type A test history supports
this conclusion.
The integrity of the containment is subject to two types of
failure mechanisms that can be categorized as: (1) Activity based,
and; (2) time based. Activity based failure mechanisms are defined
as degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment combined with the containment inspections performed in
accordance with ASME [American Society of Mechanical Engineers]
Section XI, the Maintenance Rule, and TS requirements serve to
provide a high degree of assurance that the containment would not
degrade in a manner that is detectable only by a Type A test. Based
on the above, the proposed extensions do not significantly increase
the consequences of an accident previously evaluated.
The proposed amendment also deletes exceptions previously
granted to allow one-time extensions of the ILRT test frequency for
both Units 1 and 2. These exceptions were for activities that have
already taken place so their deletion is solely an administrative
action that has no effect on any component and no impact on how the
units are operated.
Therefore, the proposed change does not result in 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 amendment to the TS involves the extension of the
CPNPP, Unit 1 and 2 Type A containment test interval to 15 years and
the extension of the Type C test interval to 75 months. The
containment and the testing requirements to periodically demonstrate
the integrity of the containment exist to ensure the plant's ability
to mitigate the consequences of an accident do not involve any
accident precursors or initiators. The proposed change does not
involve a physical change to the plant (i.e., no new or different
type of equipment will be installed) or a change to the manner in
which the plant is operated or controlled.
The proposed amendment also deletes exceptions previously
granted to allow one-time extensions of the ILRT test frequency for
both Units 1 and 2. These exceptions were for activities that would
have already taken place by the time this amendment is approved;
therefore, their deletion is solely an administrative action that
does not result in any change in how the units are operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment to TS 5.5.16 involves the extension of
the CPNPP, Units 1 and 2 Type A containment test interval to 15
years and the extension of the Type C test interval to 75 months for
selected components. This amendment does not alter the manner in
which safety limits, limiting safety system set points, or limiting
conditions for operation are determined. The specific requirements
and conditions of the TS Containment Leak Rate Testing Program exist
to ensure that the degree of containment structural integrity and
leak-tightness that is considered in the plant safety analysis is
maintained. The overall containment leak rate limit specified by TS
is maintained.
The proposed change involves only the extension of the interval
between Type A containment leak rate tests and Type C tests for
CPNPP, Units 1 and 2. The proposed surveillance interval extension
is bounded by the 15-year ILRT Interval and the 75-month Type C test
interval currently authorized within NEI 94-01, Revision 3-A.
Industry experience supports the conclusion that Type B and C
testing detects a large percentage of containment leakage paths and
that the percentage of containment leakage paths that are detected
only by Type A testing is small. The containment inspections
performed in accordance with ASME Section Xl, TS and the Maintenance
Rule serve to provide a high degree of assurance that the
containment would not degrade in a manner that is detectable only by
Type A testing. The combination of these factors ensures that the
margin of safety in the plant safety analysis is maintained. The
design, operation, testing methods and acceptance criteria for Type
A, B, and C containment leakage tests specified in applicable codes
and standards would continue to be met, with the acceptance of this
proposed change, since these are not affected by changes to the Type
A and Type C test intervals.
The proposed amendment also deletes exceptions previously
granted to allow one-time extensions of the ILRT test frequency for
both Units 1 and 2. These exceptions were for activities that would
have already taken place by the time this amendment is approved;
therefore, their deletion is solely an administrative action and
does not change how the units are operated and maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: December 4, 2014. A publicly-available
version is in ADAMs under Accession No. ML14339A637.
Description of amendment request: The proposed change would amend
Combined License (COL) Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station (VCSNS) Units 2 and 3 by changing the structure and
layout of various areas of the annex building. The proposed amendment
requires changes to the Updated Final Safety Analysis Report (UFSAR) in
the form of departures from the incorporated plant-
[[Page 17094]]
specific Design Control Document (DCD) Tier 2 information and involves
changes to related plant-specific Tier 2* and Tier 1 information, with
corresponding changes to the associated COL Appendix C information.
Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Electric Company's Advanced Passive
1000 DCD, the licensee also requested an exemption from the
requirements of the Generic DCD Tier 1 in accordance with 10 CFR
52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed additions of a new nonsafety-related battery,
battery room and battery equipment room, the room height increase,
the floor thickness changes, the relocation of a non-structural
internal wall, and the associated wall, room and corridor changes
within the annex building do not adversely affect the fire loading
analysis durations of the affected fire zones and areas (i.e., the
calculated fire durations remain less than their design values).
Thus, the fire loads analysis is not adversely affected (i.e.,
analysis results remain acceptable). The safe shutdown fire analysis
is not affected. The proposed changes to the structural
configuration, including anticipated equipment loading, room height,
and floor thickness are accounted for in the updated structural
configuration model that was used to analyze the Annex Building for
safe shutdown earthquake (SSE) and other design loads and load
combinations, thus the structural analysis is not adversely
affected. The structural analysis description and results in the
UFSAR are unchanged. The relocated internal Annex Building wall is
non-structural, thus this change does not affect the structural
analyses for the Annex Building. The proposed changes do not involve
any accident initiating event or component failure, thus the
probabilities of the accidents previously evaluated are not
affected. The rooms affected by the proposed changes do not contain
or interface with safety-related equipment, thus the proposed
changes would not affect any safety-related equipment or accident
mitigating function. The radioactive material source terms and
release paths used in the safety analyses are unchanged, thus the
radiological releases in the accident analyses are not affected.
With the conversion of an annex building room to a battery room,
the building volume serviced by nuclear island nonradioactive
ventilation system decreases by approximate five percent. This
reduced volume is used in the post-accident main control room dose
portion of the UFSAR LOCA radiological analysis. However, the volume
decrease is not sufficient to change the calculated main control
room dose reported in the UFSAR, and control room habitability is
not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed additions of a new nonsafety-related battery,
battery room and battery equipment room, the room height increase,
the floor thickness changes, the relocation of a non-structural
internal wall, and their associated wall, room and corridor changes
do not change fire barrier performance, and the fire loading
analyses results remain acceptable. The room height and floor
thickness changes are consistent with the annex building
configuration used in the building's structural analysis. The
relocated internal wall is non-structural, thus the structural
analyses for the annex building are not affected. The affected rooms
and associated equipment do not interface with components that
contain radioactive material. The affected rooms do not contain
equipment whose failure could initiate an accident. The proposed
changes do not create a new fault or sequence of events that could
result in a radioactive material release.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed additions of a new nonsafety-related battery,
battery room and battery equipment room, the room height increase,
the floor thickness changes, the relocation of a non-structural
internal wall, and their associated wall, room and corridor changes
do not change the fire barrier performance of the affected fire
areas. The affected rooms do not contain safety-related equipment,
and the safe shutdown fire analysis is not affected. Because the
proposed change does not alter compliance with the construction
codes to which the annex building is designed and constructed, the
proposed changes to the structural configuration, including
anticipated equipment loading, room height, and floor thickness do
not adversely affect the safety margins associated with the seismic
Category II structural capability of the annex building.
The floor areas and amounts of combustible material loads in
affected fire zones and areas do not significantly change, such that
their fire duration times remain within their two-hour design value,
thus the safety margins associated with the fire loads analysis are
not affected.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, thus no margin of
safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart.
South Carolina Electric and Gas Company, Docket Nos.: 52-027 and 52-
028, Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: February 10, 2015. A publicly-available
version is in ADAMS under Accession No. ML15041A698.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station (VCSNS) Units 2 and 3 by revising Tier 2* information
contained within the Human Factors Engineering Design Verification,
Task Support Verification and Integrated System Validation plans. These
documents are incorporated by reference into the VCSNS Units 2 and 3
Updated Final Safety Analysis Report and will additionally require
changes to be made to affected Tier 2 information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment includes changes to Integrated System
Validation (ISV) activities, which are performed on the AP1000 plant
simulator to validate the adequacy of the AP1000 human systems
interface design and confirm that it meets human factors engineering
principles. The proposed changes involve administrative details
related to performance of the ISV, and no plant hardware or
equipment is affected whose failure could initiate an accident, or
that interfaces with a component that could initiate an accident, or
that contains radioactive material. Therefore, these changes have no
effect on any accident initiator in the Updated Final Safety
Analysis Report (UFSAR), nor do they affect the radioactive material
releases in the UFSAR accident analysis.
[[Page 17095]]
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment includes changes to ISV activities, which
are performed on the AP1000 plant simulator to validate the adequacy
of the AP1000 human system interface design and confirm that it
meets human factors engineering principles. The proposed changes
involve administrative details related to performance of the ISV,
and no plant hardware or equipment is affected whose failure could
initiate an accident, or that interfaces with a component that could
initiate an accident, or that contains radioactive material.
Although the ISV may identify a need to initiate changes to add,
modify, or remove plant structures, systems, or components, these
changes will not be made directly as part of the ISV.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment includes changes to ISV activities, which
are performed on the AP1000 plant simulator to validate the adequacy
of the AP1000 human system interface design and confirm that it
meets human factors engineering principles. The proposed changes
involve administrative details related to performance of the ISV,
and do not affect any safety-related equipment, design code
compliance, design function, design analysis, safety analysis input
or result, or design/safety margin. No safety analysis or design
basis acceptance limit/criterion is challenged or exceeded by the
proposed changes, thus no margin of safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-
366, Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: October 10, 2014. A publicly-available
version is in ADAMS under Accession No. ML14288A226.
Description of amendment request: The licensee requested 21
revisions to the Technical Specifications. The licensee states the
changes were chosen to increase the consistency between the Hatch
Technical Specifications, the Improved Standard Technical
Specifications, and the Technical Specifications of other plants in the
Southern Nuclear Operating Company fleet. A list of the requested
revisions is included in Enclosure 1 of the application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration for each of the 24 changes requested, which is presented
below:
2.1 TSTF-30-A, Revision 3, ``Extend the Completion Time for Inoperable
Isolation Valve to a Closed System to 72 Hours.''
Specification 3.6.1.3, ``Primary Containment Isolation Valves
(PCIVs),'' Action C, TS page 3.6-9, is revised to provide a 72 hour
Completion Time for penetration flow paths with one inoperable PCIV
with a closed system.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the Completion Time to isolate an
inoperable primary containment isolation valve (PCIV) from 4 hours
to 72 hours when the PCIV is associated with a closed system. The
PCIVs are not an initiator of any accident previously evaluated. The
consequences of a previously evaluated accident during the extended
Completion Time are the same as the consequences 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 extends the Completion Time to isolate an
inoperable primary containment isolation valve (PCIV) from 4 hours
to 72 hours when the PCIV is associated with a closed system. The
PCIVs serve to mitigate the potential for radioactive release from
the primary containment following an accident. The design and
response of the PCIVs to an accident are not affected by this
change. The revised Completion Time is appropriate given the
isolation capability of the closed system.
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.
2.2 TSTF-45-A, Revision 2, ``Exempt Verification of CIVs that are
Locked, Sealed or Otherwise Secured''
The proposed change revises SRs 3.6.1.3.2 and 3.6.1.3.3 in
Specification 3.6.1.3, ``Primary Containment Isolation Valves
(PCIVs),'' to exempt manual PCIVs and blind flanges which are
locked, sealed, or otherwise secured in position from position
verification requirements. The proposed change also revises SR
3.6.4.2.1 in Specification 3.6.4.2, ``Secondary Containment
Isolation Valves (SCIVs),'' to exempt manual SCIVs and blind flanges
which are locked, sealed, or otherwise secured in position from
position verification requirements.
Signification Hazards Consideration: SNC has evaluated whether
or not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change exempts manual primary containment isolation
valves and blind flanges located inside and outside of containment,
and manual secondary containment isolation valves and blind flanges,
that are locked, sealed, or otherwise secured in position from the
periodic verification of valve position required by Surveillance
Requirements 3.6.1.3.2, 3.6.1.3.3, and 3.6.4.2.1. The exempted
valves and devices are verified to be in the correct position upon
being locked, sealed, or secured. Because the valves and devices 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.
[[Page 17096]]
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 exempts manual primary containment isolation
valves and blind flanges located inside and outside of containment,
and manual secondary containment isolation valves and blind flanges,
that are locked, sealed, or otherwise secured in position from the
periodic verification of valve position required by Surveillance
Requirements 3.6.1.3.2, 3.6.1.3.3, and 3.6.4.2.1. These valves and
devices are administratively controlled and their operation is a
non-routine event. The position of a locked, sealed or secured blind
flange or valve is verified at the time it is locked, sealed or
secured, and any changes to their position is performed under
administrative controls. Industry experience has shown that these
valves are generally found to be in the correct position. Since the
change impacts only the frequency of verification for blind flange
and valve position, the proposed change will provide a similar level
of assurance of correct position as the current frequency of
verification.
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.
2.3 TSTF-46-A, Revision 1, ``Clarify the CIV Surveillance to Apply Only
to Automatic Isolation Valves''
The proposed change modifies SR 3.6.1.3.5 in Specification
3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' and SR
3.6.4.2.2, in Specification 3.6.4.2, ``Secondary Containment
Isolation Valves (SCIVs),'' including their associated Bases, to
delete the requirement to verify the isolation time of ``each power
operated'' containment isolation valve and only require verification
of each ``power operated automatic isolation valve.''
Signification Hazards Consideration: SNC has evaluated whether
or not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the requirements in Technical
Specification Surveillance Requirements (SRs) 3.6.1.3.5 and
3.6.4.2.2, and their associated Bases, to delete the requirement to
verify the isolation time of ``each power operated'' PCIV and SCIV
and only require verification of closure time for each ``automatic
power operated isolation valve.'' The closure times for PCIVs and
SCIVs 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 PCIVs and SCIVs are used to respond to
accidents previously evaluated. Power operated PCIVs and SCIVs 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 PCIVs and SCIVs provide plant protection or introduce any
new or different operational conditions. Periodic verification that
the closure times for PCIVs and SCIVs that receive an automatic
closure signal are within the limits established by the accident
analysis will continue to be performed under SRs 3.6.1.3.5 and
3.6.4.2.2. 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 PCIVs and
SCIVs that receive an automatic isolation signal are within the
scope of SRs 3.6.1.3.5 and 3.6.4.2.2. The proposed change does not
result in a change in the manner in which the PCIVs and SCIVs
provide plant protection. Periodic verification that closure times
for PCIVs and SCIVs 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.
2.4 TSTF-222-A, Revision 1, ``Control Rod Scram Time Testing''
Specification 3.1.4, ``Control Rod Scram Times,'' SRs 3.1.4.1
and 3.1.4.4, are revised to only require scram time testing of
control rods that are in an affected core cell. The SR 3.1.4.1
Frequency ``Prior to exceeding 40% RTP after fuel movement within
the reactor vessel,'' is eliminated and a new Frequency is added to
SR 3.1.4.4 which states, ``Prior to exceeding 40% RTP after fuel
movement within the affected core cell.''
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change clarifies the intent of Surveillance testing
in Specification 3.1.4, ``Control Rod Scram Times.'' The existing
Specification wording requires control rod scram time testing of all
control rods whenever fuel is moved within the reactor pressure
vessel, even though the Technical Specification Bases state that
control rod scram time testing is only required in the affected core
cells. The Frequency of Surveillances 3.1.4.1 and 3.1.4.4 are
revised to implement the Bases statement in the Specifications. The
proposed change does not affect any plant equipment, test methods,
or plant operation, and are not initiators of any analyzed accident
sequence. The control rods will continue to perform their function
as designed. Operation in accordance with the proposed Technical
Specifications will ensure that all analyzed accidents will continue
to be mitigated as previously analyzed.
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
[[Page 17097]]
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 clarifies the intent of Surveillance testing
in Specification 3.1.4, ``Control Rod Scram Times.'' The existing
Specification wording requires control rod scram time testing of all
control rods whenever fuel is moved within the reactor pressure
vessel, even though the Technical Specification Bases state that the
control rod scam time testing is only required in the affected core
cells. The proposed change will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. Control rod scram time testing will be performed following
any fuel movement that could affect the scram 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.
2.5 TSTF-264-A, Revision 0, ``3.3.9 and 3.3.10--Delete Flux Monitors
Specific Overlap Requirement SRs''
The proposed change revises Specification 3.3.1.1, ``RPS
Instrumentation,'' by deleting Surveillances 3.3.1.1.6 and
3.3.1.1.7, which verify the overlap between the source range monitor
(SRM) and the intermediate range monitor (IRM), and between the IRM
and the average power range monitor (APRM).
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates two Surveillances Requirements
(SRs) (SRs 3.3.1.1.6 and 3.3.1.1.7) which verify the overlap between
the source range monitor (SRM) and intermediate range monitor (IRM)
and between the IRM and the average power range monitor (APRM). The
testing requirement is incorporated in the existing Channel Check
Surveillance (SR 3.3.1.1.1). The proposed change does not affect any
plant equipment, test methods, or plant operation, and are not
initiators of any analyzed accident sequence. The SRM, IRM, and APRM
will continue to perform their function as designed. Operation in
accordance with the proposed Technical Specifications will ensure
that all analyzed accidents will continue to be mitigated as
previously analyzed.
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 SRs 3.3.1.1.6 and 3.3.1.1.7 which
verify the overlap between the SRM and IRM and between the IRM and
the APRM. The testing requirement is incorporated in the existing
Channel Check Surveillance (SR 3.3.1.1.1). The proposed change will
not affect the operation of plant equipment or the function of any
equipment assumed in the accident analysis. Instrument channel
overlap will continue to be verified under the existing Channel
Check surveillance.
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.
2.6 TSTF-269-A, Revision 2, ``Allow Administrative Means of Position
Verification for Locked or Sealed Valves''
The proposed change modifies Specification 3.6.1.3, ``Primary
Containment Isolation Valves,'' and Specification 3.6.4.2,
``Secondary Containment Isolation Valves.'' The specifications
require penetrations with an inoperable isolation valve to be
isolated and periodically verified to be isolated. A Note is added
to Specification 3.6.1.3, Actions A and C, and Specification
3.6.4.2, Action A, to allow isolation devices that are locked,
sealed, or otherwise secured to be verified by use of administrative
means.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies Specification 3.6.1.3, ``Primary
Containment Isolation Valves,'' and Specification 3.6.4.2,
``Secondary Containment Isolation Valves.'' The specifications
require penetrations with an inoperable isolation valve to be
isolated and periodically verified to be isolated. A Note is added
to Specification 3.6.1.3, Actions A and C, and Specification
3.6.4.2, Action A, to allow isolation devices that are locked,
sealed, or otherwise secured to be verified by use of administrative
means. The proposed change does not affect any plant equipment, test
methods, or plant operation, and are not initiators of any analyzed
accident sequence. The inoperable containment penetrations will
continue to be isolated, and hence perform their isolation function.
Operation in accordance with the proposed Technical Specifications
will ensure that all analyzed accidents will continue to be
mitigated as previously analyzed.
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 will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. The primary and secondary containment isolation valves
will continue to be operable or will be isolated as required by the
existing specifications.
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.
2.7 TSTF-273-A, Revision 2, ``Safety Function Determination Program
Clarifications''
The proposed Technical Specification (TS) changes add
explanatory text to the Bases for limiting condition for operation
(LCO) 3.0.6 clarifying 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. Explanatory text is also added to
the programmatic description of the Safety Function Determination
Program (SFDP) in Specification 5.5.12 to provide clarification of
these same issues.
Signification Hazards Consideration: SNC has evaluated whether
or not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the
[[Page 17098]]
three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification (TS) changes add
explanatory text to the programmatic description of the Safety
Function Determination Program (SFDP) in Specification 5.5.10 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 limiting condition for operations (LCO) 3.0.6 are 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 5.5.10 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.
2.8 TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode Restriction
Notes''
The proposed change revises several Specification 3.8.1, ``AC
Sources--Operating,'' Surveillance Notes to allow full or partial
performance of the SRs to re-establish Operability provided an
assessment determines the safety of the plant is maintained or
enhanced. These Surveillances currently have Notes prohibiting their
performance in Modes 1 or 2, or in Modes 1, 2, or 3.
SR 3.8.1.6 (ISTS SR 3.8.1.8), which tests the transfer of
Alternating (AC) sources from normal to alternate offsite circuits,
contains a Note prohibiting performance in Mode 1 or 2. The Note is
modified to state that performance is normally prohibited in Mode 1
or 2 but may be performed to re-establish Operability provided an
assessment determines the safety of the plant is maintained or
enhanced.
SR 3.8.1.7 (ISTS SR 3.8.1.9), which tests the ability of the
emergency diesel generator (DG) to reject a load greater than or
equal to its associated single largest post-accident load, contains
a Note prohibiting performance in Mode 1 or 2. An exception is
provided for the swing DG. The Note is modified to state that
performance is normally prohibited in Mode 1 or 2 but may be
performed to re- establish Operability provided an assessment
determines the safety of the plant is maintained or enhanced.
SR 3.8.1.8 (ISTS SR 3.8.1.10), which tests emergency DG
operation following a load rejection of greater than or equal to
2775 kW, contains a Note prohibiting performance in Mode 1 or 2. The
Note is modified to state that performance is normally prohibited in
Mode 1 or 2 but portions of the SR may be performed to re- establish
Operability provided an assessment determines the safety of the
plant is maintained or enhanced.
SR 3.8.1.9 (ISTS SR 3.8.1.11), which tests the response to a
loss of offsite power signal, contains a Note prohibiting
performance in Mode 1, 2, or 3. The Note is modified to state that
performance is normally prohibited in Mode 1, 2, or 3, but portions
of the SR may be performed to re-establish Operability provided an
assessment determines the safety of the plant is maintained or
enhanced.
SR 3.8.1.10 (ISTS SR 3.8.1.12), which tests response to an
Emergency Core Cooling System (ECCS) initiation signal, contains a
Note prohibiting performance in Mode 1 or 2. The Note is modified to
state that performance is normally prohibited in Mode 1 or 2, but
the SR may be performed to re-establish Operability provided an
assessment determines the safety of the plant is maintained or
enhanced.
SR 3.8.1.11 (ISTS SR 3.8.1.13), which tests that each DGs
automatic trips are bypassed on a loss of voltage signal concurrent
with an ECCS initiation signal, contains a Note prohibiting
performance in Mode 1, 2, or 3. The Note is modified to state that
performance is normally prohibited in Mode 1, 2, or 3, but the SR
may be performed to re-establish Operability provided an assessment
determines the safety of the plant is maintained or enhanced.
SR 3.8.1.12 (ISTS SR 3.8.1.14), which performs a 24 hour loaded
test run of the DG, contains a Note prohibiting performance in Mode
1 or 2. The Note is modified to state that performance is normally
prohibited in Mode 1 or 2, but the SR may be performed to re-
establish Operability provided an assessment determines the safety
of the plant is maintained or enhanced.
SR 3.8.1.14 (ISTS SR 3.8.1.16), which verifies transfer from DG
to offsite power, contains a Note prohibiting performance in Mode 1,
2, or 3. The Note is modified to state that performance is normally
prohibited in Mode 1, 2, or 3, but portions of the SR may be
performed to re-establish Operability provided an assessment
determines the safety of the plant is maintained or enhanced.
SR 3.8.1.15 (ISTS SR 3.8.1.17), which verifies than a DG
operating in test mode will return to ready-to-load condition and
energize the emergency load from offsite power on receipt of an ECCS
initiation signal, contains a Note prohibiting performance in Mode
1, 2, or 3. The Note is modified to state that performance is
normally prohibited in Mode 1, 2, or 3, but portions of the SR may
be performed to re-establish Operability provided an assessment
determines the safety of the plant is maintained or enhanced.
SR 3.8.1.16 (ISTS SR 3.8.1.18), which verifies the interval
between each sequenced load, contains a Note prohibiting performance
in Mode 1, 2, or 3. The Note is modified to state that performance
is normally prohibited in Mode 1, 2, or 3, but the SR may be
performed to re-establish Operability provided an assessment
determines the safety of the plant is maintained or enhanced.
SR 3.8.1.17 (ISTS SR 3.8.1.19), which verifies the response to a
loss of offsite power signal and Engineered Safety Features (ESF)
actuation signal, contains a Note prohibiting performance in Mode 1,
2, or 3. The Note is modified to state that performance is normally
prohibited in Mode 1, 2, or 3, but portions of the SR may be
performed to re-establish Operability provided an assessment
determines the safety of the plant is maintained or enhanced.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
[[Page 17099]]
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 Mode restriction Notes on eleven
emergency diesel generator (DG) Surveillances to allow performance
of the Surveillance in whole or in part to re-establish emergency DG
Operability. The emergency DGs and their associated emergency loads
are accident mitigating features, and are not an initiator of any
accident previously evaluated. As a result the probability of any
accident previously evaluated is not increased. The proposed change
allows Surveillance testing to be performed in whole or in part to
re-establish Operability of an emergency DG. The consequences of an
accident previously evaluated during the period that the emergency
DG is being tested to re-establish Operability are no different from
the consequences of an accident previously evaluated while the
emergency DG is inoperable. As a result, the consequences of any
accident previously evaluated are not increased.
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 purpose of Surveillances is to verify that equipment is
capable of performing its assumed safety function. The proposed
change will only allow the performance of the Surveillances to re-
establish Operability and the proposed changes may not be used to
remove an emergency DG from service. The proposed changes also
require an assessment to verify that plant safety will be maintained
or enhanced by performance of the Surveillance in the normally
prohibited Modes.
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.
2.9 TSTF-284-A, Revision 3, ``Add `Met vs. Perform' to Technical
Specification 1.4, Frequency''
The change inserts a discussion paragraph into Specification
1.4, and two new examples are added to facilitate the use and
application of SR Notes that utilize the terms ``met'' and
``perform.''
Signification Hazards Consideration: SNC has evaluated whether
or not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously 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 Surveillance Requirement (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
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 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.
2.10 TSTF-295-A, Revision 0, ``Modify Note 2 to Actions of PAM Table to
Separate Condition Entry for Each Penetration''
Specification 3.3.3.1, ``Post Accident Monitoring (PAM)
Instrumentation,'' Function 6, is renamed from ``Primary Containment
Isolation Valve Position'' to ``Penetration Flow Path Primary
Containment Isolation Valve Position.''
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change clarifies the separate condition entry Note
in Specification 3.3.3.1, ``Post Accident Monitoring (PAM)
Instrumentation,'' for Function 6, ``Primary Containment Isolation
Valve Position,'' and Function 9, ``Suppression Pool Water
Temperature.'' The proposed change does not affect any plant
equipment, test methods, or plant operation, and are not initiators
of any analyzed accident sequence. The actions taken for inoperable
PAM channels are not changed. Operation in accordance with the
proposed Technical Specifications will ensure that all analyzed
accidents will continue to be mitigated as previously analyzed.
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
[[Page 17100]]
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 will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. The PAM channels will continue to be operable or the
existing, appropriate actions will be followed.
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.
2.11 TSTF-306-A, Revision 2, ``Add Action to LCO 3.3.6.1 to Give Option
to Isolate the Penetration''
The proposed change revises Specification 3.3.6.1, ``Primary
Containment Isolation Instrumentation.'' An Actions Note is added
allowing penetration flow paths to be unisolated intermittently
under administrative controls. The traversing incore probe (TIP)
isolation system is also segregated into a separate Function,
allowing 12 hours to place the channel in trip and 24 hours to
isolate the penetration. A new Condition G is added for the new TIP
isolation system Function. Condition G is referenced from Required
Action C.1 when Conditions A or B are not met. The subsequent
Actions are renumbered.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Specification 3.3.6.1, ``Primary
Containment Isolation Instrumentation.'' An Actions Note is added
allowing penetration flow paths to be unisolated intermittently
under administrative controls. The traversing incore probe (TIP)
isolation system is segregated into a separate Function, allowing 12
hours to place the channel in trip and 24 hours to isolate the
penetration. A new Action G is added which is referenced by the new
TIP isolation system Function. The subsequent Actions are
renumbered. The proposed change does not affect any plant equipment,
test methods, or plant operation, and are not initiators of any
analyzed accident sequence. The allowance to unisolate a penetration
flow path will not have a significant effect on mitigation of any
accident previously evaluated because the penetration flow path can
be isolated, if needed, by a dedicated operator. The option to
isolate a TIP System penetration will ensure the penetration will
perform as assumed in the accident analysis. Operation in accordance
with the proposed Technical Specifications will ensure that all
analyzed accidents will continue to be mitigated as previously
analyzed.
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 will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. The allowance to unisolate a penetration flow path will
not have a significant effect on a margin of safety because the
penetration flow path can be isolated manually, if needed. The
option to isolate a TIP System penetration will ensure the
penetration will perform 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.
2.12 TSTF-308-A, Revision 1, ``Determination of Cumulative and
Projected Dose Contributions in RECP''
The proposed change revises Specification 5.5.4, ``Radioactive
Effluent Controls Program,'' paragraph e, to describe the original
intent of the dose projections.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously 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.
2.13 TSTF-318-A, Revision 0, ``Revise 3.5.1 for One LPCI Pump
Inoperable in Each of Two ECCS Divisions''
The proposed change adds a provision to Condition A of
Specification 3.5.1, ``ECCS--Operating,'' to allow one Low Pressure
Coolant Injection (LPCI) pump to be inoperable in each subsystem for
a period of seven days.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change adds a provision to Condition A of Technical
Specification (TS) 3.5.1 to allow one Low Pressure Coolant Injection
(LPCI) pump to be inoperable in each subsystem for a period of seven
days. The change to allow one LPCI pump to be inoperable in both
subsystems is more reliable than what is currently allowed by
Condition A, which requires entry into
[[Page 17101]]
shutdown limiting condition for operation (LCO) 3.0.3 under these
conditions. The LPCI mode of the Residual Heat Removal system is not
assumed to be initiator of any analyzed event sequence. The
consequences of an accident previously evaluated under the proposed
allowance are no different than the consequences under the existing
requirements.
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 a provision to Condition A of Technical
Specification TS 3.5.1 to allow one LPCI pump to be inoperable in
each subsystem for a period of seven days. The change to allow one
LPCI pump to be inoperable in both subsystems is more reliable than
what is currently allowed by Condition A, which requires entry into
shutdown LCO 3.0.3 under these conditions. The proposed change does
not affect any safety analysis assumptions.
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.
2.14 TSTF-322-A, Revision 2, ``Secondary Containment and Shield
Building Boundary Integrity SRs'
The proposed change revises Specification 3.6.4.1, ``Secondary
Containment,'' SRs 3.6.4.1.3 and 3.6.4.1.4 to clarify the intent of
the Surveillances.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Specification 3.6.4.1, ``Secondary
Containment,'' Surveillance Requirements (SRs) 3.6.4.1.3 and
3.6.4.1.4 to clarify the intent of the Surveillances. The secondary
containment and the standby gas treatment (SGT) system are not
initiators of any accident previously evaluated. Operation in
accordance with the proposed Technical Specifications will ensure
that all analyzed accidents will continue to be mitigated as
previously analyzed.
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 is an clarification of the intent of the
surveillances to ensure that the secondary containment is not
inappropriately declared inoperable when a SGT subsystem is
inoperable. The safety functions of the secondary containment and
the SGT system are not affected. This change is a correction that
ensures that the intent of the secondary containment surveillances
is clear.
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.
2.15 TSTF-323-A, Revision 0, ``EFCV Completion Time to 72 hours''
The proposed change revises Specification 3.6.1.3, ``Primary
Containment Isolation Valves,'' Action C, to provide a 72 hour
Completion Time instead of a 12 hour Completion Time to isolate an
inoperable excess flow check valve (EFCV).
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Specification 3.6.1.3, ``Primary
Containment Isolation Valves,'' Action C, to provide a 72 hour
Completion Time instead of a 12 hour Completion Time to isolate an
inoperable excess flow check valve (EFCV). The primary containment
isolation valves (PCIVs) are not an initiator of any accident
previously evaluated. The consequences of a previously evaluated
accident during the extended Completion Time are the same as the
consequences 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 extends the Completion Time to isolate an
inoperable primary containment penetration equipped with an excess
flow check valve from 12 hours to 72 hours. The PCIVs serve to
mitigate the potential for radioactive release from the primary
containment following an accident. The design and response of the
PCIVs to an accident are not affected by this change. The revised
Completion Time is appropriate given the EFCVs are on penetrations
that have been found to have acceptable barrier(s) in the event that
the single isolation valve fails.
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.
2.16 TSTF-374-A, Revision 0, ``Revision to TS 5.5.13 and Associated TS
Bases for Diesel Fuel Oil''
The proposed change revises Specification 5.5.9, ``Diesel Fuel
Oil Testing Program,'' to remove references to the specific American
Society for Testing and Materials (ASTM) Standard from the
Administrative Controls Section of TS, and places them in a
licensee-controlled document. Also, alternate criteria are added to
establish the acceptability of new fuel oil for use prior to and
following the addition to storage tanks.
Signification Hazards Consideration: SNC has evaluated whether
or not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 17102]]
Response: No.
The proposed changes remove the references to specific ASTM
standards from the Administrative Controls Section of the Technical
Specifications (TS) and place them in a licensee controlled
document. Requirements to perform testing in accordance with the
applicable ASTM standards is retained in the TS as are requirements
to perform testing of both new and stored diesel fuel oil. Future
changes to the licensee controlled document will be evaluated
pursuant to the requirements of 10 CFR 50.59 to ensure that these
changes do not result in more than a minimal increase in the
probability or consequences of an accident previously evaluated. In
addition, tests used to establish the acceptability of new fuel oil
for use prior to and following the addition to storage tanks has
been expanded to recognize more rigorous testing of water and
sediment content. Relocating the specific ASTM standard references
from the TS to a licensee controlled document and allowing a water
and sediment content test to be performed to establish the
acceptability of new fuel oil will not affect nor degrade the
ability of the emergency diesel generators (EDGs) to perform their
specified safety function. Fuel oil quality will continue to be
tested and maintained to ASTM requirements. Diesel fuel oil testing
is not an initiator of any accident previously evaluated, and the
proposed changes do not adversely affect any accident initiators or
precursors, or alter design assumptions, conditions, and
configuration of the facility, or the manner in which the plant is
operated. The proposed changes do not adversely affect the ability
of structures, systems, and components to perform their intended
safety function to mitigate the consequences of an initiating event
within the assumed acceptance limits.
Therefore, the proposed changes do 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 changes remove the references to specific ASTM
standards from the Administrative Controls Section of TS and place
them in a licensee controlled document. In addition, the tests used
to establish the acceptability of new fuel oil for use prior to and
following the addition to storage tanks has been expanded to allow a
water and sediment content test to be performed to establish the
acceptability of new fuel oil. 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. The requirements retained in the TS will continue to
require testing of new and stored diesel fuel oil to ensure the
proper functioning of the EDGs.
Therefore, the 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 remove the references to specific ASTM
standards from the Administrative Controls Section of TS and place
them in a licensee controlled document. Instituting the proposed
changes will continue to ensure the use of applicable ASTM standards
to evaluate the changes to the licensee-controlled document are
performed in accordance with the provisions of 10 CFR 50.59. This
approach provides an effective level of regulatory control and
ensures that diesel fuel oil testing is conducted such that there is
no significant reduction in a margin of safety. The margin of safety
provided by the EDGs is unaffected by the proposed changes since TS
requirements will continue to ensure fuel oil is of the appropriate
quality. The proposed changes provide the flexibility needed to
improve fuel oil sampling and analysis methodologies while
maintaining sufficient controls to preserve the current margins 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.
2.17 TSTF-400-A, Revision 1, ``Clarify SR on Bypass of DG Automatic
Trips''
The proposed change revises Specification 3.8.1, ``AC Sources--
Operating,'' Surveillance 3.8.1.11, to clarify that the intent of
the SR is to test the non-critical emergency DG automatic trips.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change clarifies the purpose of Surveillance Requirement
(SR) 3.8.1.11, which is to verify that non-critical automatic
emergency diesel generator (DG) trips are bypassed in an accident.
The non-critical automatic DG trips and their bypasses are not
initiators of any accident previously evaluated. Therefore, the
probability of any accident is not significantly increased.
Additionally, the function of the emergency DG in mitigating
accidents is not changed. The revised SR continues to ensure the
emergency DG will operate as assumed in the 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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change clarifies the purpose of SR 3.8.1.11, which is to
verify that non-critical automatic emergency DG trips are bypassed
in an accident. The proposed change does not involve a physical
alteration of the plant (no new or different type of equipment will
be installed), or a change in the methods governing normal plant
operation. Thus, this change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change clarifies the purpose of SR 3.8.1.11, which is to
verify that non-critical automatic DG trips are bypassed in an
accident. This change clarifies the purpose of the SR, which is to
verify that the emergency DG is capable of performing the assumed
safety function. The safety function of the emergency DG is
unaffected, so the change does not affect the margin of safety.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based on the above, SNC concludes that the proposed change
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
2.18 TSTF-439-A, Revision 2, ``Eliminate Second Completion Times
Limiting Time From Discovery of Failure To Meet an LCO''
Specifications 3.1.7, ``Standby Liquid Control (SLC) System;''
3.6.4.3, ``Standby Gas Treatment (SGT) System;'' 3.8.1, ``AC
Sources--Operating;'' and 3.8.7, ``Distribution Systems--
Operating,'' contain Required Actions with a second Completion Time
to establish a limit on the maximum time allowed for any combination
of Conditions that result in a single continuous failure to meet the
LCO. These Completion Times (henceforth referred to as ``second
Completion Times'') are joined by an ``AND'' logical connector to
the Condition-specific Completion Time and state ``X days from
discovery of failure to meet the LCO'' (where ``X'' varies by
specification). The proposed change deletes these second Completion
Times from the affected Required Actions. It also revises ISTS
Example 1.3-3 to remove the discussion of second Completion Times
and to revise the discussion in that Example to state that
alternating between Conditions in such a manner that operation could
continue indefinitely without restoring systems to meet the LCO is
inconsistent with the basis of the Completion Times and is
inappropriate. Therefore, the licensee shall have administrative
controls to limit the maximum time allowed for any combination of
Conditions that result in a single contiguous occurrence of failing
to meet the LCO.
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
[[Page 17103]]
The proposed change eliminates certain Completion Times from the
Technical Specifications. Completion Times are not an initiator to
any accident previously evaluated. As a result, the probability of
an accident previously evaluated is not affected. The consequences
of an accident during the remaining Completion Time are no different
than the consequences of the same accident during the removed
Completion Times. As a result, the consequences of an accident
previously evaluated are not affected by this change.
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 to delete the second Completion Time does
not alter the manner in which safety limits, limiting safety system
settings or limiting conditions for operation are determined. The
safety analysis acceptance criteria are not affected by this change.
The proposed changes will not result in plant operation in a
configuration outside of 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.
2.19 TSTF-458-T, Revision 0, ``Removing Restart of Shutdown Clock for
Increasing Suppression Pool Temperature''
The proposed change revises Specification 3.6.2.1, ``Suppression
Pool Average Temperature,'' Required Actions D and E, to eliminate
redundant requirements.
Significant Hazards Consideration SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Specification 3.6.2.1, ``Suppression
Pool Average Temperature,'' Required Actions D and E, to eliminate
redundant requirements when suppression pool temperature is above
the Limiting Conditions for Operation (LCO) limit. Suppression pool
temperature is not an initiator to any accident previously
evaluated. Suppression pool temperature may affect the mitigation of
accidents previously evaluated. The proposed change reduces the time
allowed to operate with suppression pool temperature above the
limit. The consequences of an accident under the proposed change are
no different than under the current requirements.
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 3.6.2.1, ``Suppression
Pool Average Temperature,'' Required Actions D and E, to eliminate
redundant requirements when suppression pool temperature is above
the LCO limit. The proposed change reduces the time allowed to
operate with suppression pool temperature above the limit. The
proposed revision will not adversely affect the margin of safety as
it corrects the Actions to provide appropriate compensatory measures
when suppression pool temperature is greater than the limit.
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.
2.20 TSTF-464-T, Revision 0, ``Clarify the Control Rod Block
Instrumentation Required Action''
The proposed change revises Specification 3.3.2.1, Required
Action C.2.1.2 from ``Verify by administrative methods that startup
with RWM inoperable has not been performed in the last calendar
year'' to ``Verify by administrative methods that startup with RWM
inoperable has not been performed in the last 12 months.''
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises a Required Action to limit startup
with the Rod Worth Minimizer (RWM) inoperable from once per calendar
year to once per 12 months. The RWM is used to minimize the
possibility and consequences of a control rod drop accident. This
change clarifies the intent of the limitation, but does not affect
the requirement for the RWM to be operable. As, over time, the
number of startups with the RWM inoperable will not increase, the
probability of any accident previously evaluated is not
significantly increased. As the RWM is still required to be
operable, the consequences of an any accident previously evaluated
are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises a Required Action to limit startup
with the Rod Worth Minimizer inoperable from once per calendar year
to once per 12 months. 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 significant change in the methods governing
normal plant operation. 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 change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises a Required Action to limit startup
with the Rod Worth Minimizer (RWM) inoperable from once per calendar
year to once per 12 months. This clarifies the intent of the
Required Action. The number of startups with RWM inoperable is not
increased.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed change
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
2.21 ISTS Adoption #1--Revise the 5.5.7 Introductory Paragraph To Be
Consistent With the ISTS
The proposed change revises the introductory paragraph of
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),''
to be consistent with the ISTS. Specific requirements to perform
testing after
[[Page 17104]]
structural maintenance on the HEPA filter or charcoal adsorber
housing or following painting, fire or chemical release, and after
every 720 hours of operation are relocated to the licensee-
controlled program.
The existing wording states, ``The VFTP will establish the
required testing of Engineered Safety Feature (ESF) filter
ventilation systems at the frequencies specified in Regulatory Guide
1.52, Revision 2, Sections C.5.c and C.5.d, or: (1) After any
structural maintenance on the HEPA filter or charcoal adsorber
housings, (2) following painting, fire or chemical release in any
ventilation zone communicating with the system, or 3) after every
720 hours of charcoal adsorber operation.''
The proposed wording states, ``A program shall be established to
implement the following required testing of Engineered Safety
Feature (ESF) filter ventilation systems at the frequencies
specified in Regulatory Guide 1.52, Revision 2, Sections C.5.c and
C.5.d, and in accordance with Regulatory Guide 1.52, Revision 2.''
Significant Hazards Consideration: SNC has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment(s) by focusing on the three standards set forth
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the introductory paragraph of
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),''
to be consistent with the ISTS. Specific requirements to perform
testing after structural maintenance on the HEPA filter or charcoal
adsorber housing or following painting, fire or chemical release,
and after every 720 hours of operation are retained as a reference
to Regulatory Guide requirements and general requirements in
Surveillance Requirement (SR) 3.0.1. Implementation of these
requirements will be in the licensee-controlled VFTP. The VFTP will
be maintained in accordance with 10 CFR 50.59. Since any changes to
the VFTP will be evaluated under 10 CFR 50.59, no significant
increase in the probability or consequences of an accident
previously evaluated will be allowed.
Therefore, this proposed change does not represent 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 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 introductory paragraph of
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),''
to be consistent with the ISTS. The proposed change will not reduce
a margin of safety because it has no effect on any safety analysis
assumption. In addition, no requirements are being removed, but are
being replaced with references to an NRC Regulatory Guide and the
requirements of SR 3.0.1.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Inverness Center
Parkway, Birmingham, AL 35201
NRC Branch Chief: Robert J. Pascarelli.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of amendment request: December 11, 2014 (ADAMS Accession No.
ML14349A694).
Description of amendment request: The amendment would revise
Section 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' of the
Technical Specifications (TSs) by replacing the current volume
requirements with the number of continuous days the diesel generators
(DGs) are required to run. The numerical volumes will be maintained in
the licensee-controlled TSs Bases document so they may be modified
under licensee control. The resulting requirements will specify an
inventory of stored diesel fuel oil and lube oil sufficient for a 7-day
supply for each DG. This proposed amendment is consistent with NRC's
approved Technical Specifications Task Force (TSTF) Improved Standard
Technical Specifications Change Traveler TSTF-501, Revision 1,
``Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee
Control.'' The availability of this TSs improvement was announced in
the Federal Register on May 26, 2010 (75 FR 29588). The licensee also
proposed additional changes to Section 3.8.3 and Section 5.5.9,
``Diesel Fuel Oil Testing Program,'' to support other related changes
proposed by TSTF-501, Revision 1. These additional changes concern fuel
oil quality and associated surveillance requirements (SRs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to TS Section 3.8.3, Conditions A and B,
and to SR 3.8.3.1 and SR 3.8.3.2 remove the volume of diesel fuel
oil and lube oil required to support 7-day operation of each onsite
diesel generator, and the volume equivalent to a 6-day supply, from
the TS and replace them with the associated number of days. The
numerical volumes will be maintained under licensee control. The
specific volume of fuel oil equivalent to a 7 and 6-day supply is
calculated using the NRC-approved methodology described in
Regulatory Guide 1.137, Revision 1, ``Fuel-Oil Systems for Standby
Diesel Generators'' and ANSI [American National Standards
Institute]-N195 1976, ``Fuel Oil Systems for Standby Diesel-
Generators.'' The specific volume of lube oil equivalent to a 7-day
and 6-day supply is based on the diesel generator manufacturer's
consumption values for the run time of the diesel generator. Because
the requirement to maintain a 7-day supply of diesel fuel oil and
lube oil is not changed and is consistent with the assumptions in
the accident analyses, and the actions taken when the volume of fuel
oil and lube oil are less than a 6-day supply have not changed,
neither the probability nor the consequences of any accident
previously evaluated will be affected.
The addition of a new Condition D provides a required action and
completion time if new fuel oil properties are not within limits.
The new SR 3.8.3.5 requires checking for and removing water from the
7-day storage tank every 31 days. The revised Section 5.5.9 adds
testing requirements for new fuel oil to be completed prior to the
addition of the new fuel oil to the 7-day storage tank, as well as
additional testing to be completed prior or within 31 days of the
addition. These requirements are more restrictive testing
requirements and provide corrective action to be taken if the
testing limits are not met. They are taken from the current NRC
approved NUREG-1433, Revision 4, ``Standard Technical
Specifications, General Electric BWR/4 Plants.'' Improved, more
restrictive testing standards will neither change the probability or
the consequences of any accident previously evaluated be affected.
Therefore, the proposed changes do not involve a significant
increase in the
[[Page 17105]]
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 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. The change
does not alter assumptions made in the safety analysis but ensures
that the diesel generator operates as assumed in the accident
analysis. The proposed change is consistent with the safety analysis
assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes to Section 3.8.3, Conditions A and B, and
to SR 3.8.3.1 and SR 3.8.3.2 remove the numerical volume of diesel
fuel oil and lube oil required to support 7-day operation of each
onsite diesel generator, and the numerical volume equivalent to a 6-
day supply from the TS and replaces them with the associated number
of days. The numerical volumes will be maintained under licensee
control. As the bases for the existing limits on diesel fuel oil
volume and lube oil volume are not changed, no change is made to the
accident analysis assumptions and no margin of safety is reduced as
part of this change.
The new, more restrictive, testing requirements, and the
provision for corrective action to be taken if the testing limits
are not met, are taken from the current NRC approved NUREG-1433,
Revision 4, ``Standard Technical Specifications, General Electric
BWR/4 Plants.'' These changes do not revise the accident analysis
assumptions and no margin of safety is reduced as part of these
changes.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Shana R. Helton.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: November 20, 2014. A publicly-available
version is in ADAMS under Accession No. ML14330A247.
Description of amendment request: The amendment would revise the
Technical Specification (TS) requirements to address NRC Generic Letter
2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay
Heat Removal, and Containment Spray Systems,'' as described in
Technical Specification Task Force (TSTF) Traveler TSTF-523, Revision
2, ``Generic Letter 2008-01, Managing Gas Accumulation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises or adds SRs [surveillance
requirements] that require verification that the Emergency Core
Cooling System (ECCS), the Residual Heat Removal (RHR) System, and
the Containment Spray System are not rendered inoperable due to
accumulated gas and to provide allowances which permit performance
of the revised verification. Gas accumulation in the subject systems
is not an initiator of any accident previously evaluated. As a
result, the probability of any accident previously evaluated is not
significantly increased. The proposed SRs ensure that the subject
systems continue to be capable to perform their assumed safety
function and are not rendered inoperable due to gas accumulation.
Thus, the consequences of any accident previously evaluated are not
significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, the RHR System, and the Containment
Spray System are not rendered inoperable due to accumulated gas and
to provide allowances which permit performance of the revised
verification. The proposed change does not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the proposed change does not impose any new
or different requirements that could initiate an accident. The
proposed change does not alter assumptions made in the safety
analysis and is consistent with the safety analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, the RHR System, and the Containment
Spray System are not rendered inoperable due to accumulated gas and
to provide allowances which permit performance of the revised
verification. The proposed change adds new requirements to manage
gas accumulation in order to ensure the subject systems are capable
of performing their assumed safety functions. The proposed SRs are
more comprehensive than the current SRs and will ensure that the
assumptions of the safety analysis are protected. The proposed
change does not adversely affect any current plant safety margins or
the reliability of the equipment assumed in the safety analysis.
Therefore, there are no changes being made to any safety analysis
assumptions, safety limits or limiting safety system settings that
would adversely affect plant safety as a result of the proposed
change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
III. Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses and Combined Licenses,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
[[Page 17106]]
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: August 22, 2014. A publicly-available
version is in ADAMS under Accession No. ML14237A729.
Brief description of amendment request: The proposed amendment
would revise the technical specification (TS) surveillance requirement
(SR) for the ultimate heat sink (UHS) to clarify that spray pond level
is the average of the level in both ponds. The design of the ultimate
heat sink is such that it is difficult to meet the current SR when only
one standby service water (SW) pump is in operation without overflowing
a spray pond resulting in a net loss of water inventory, which may
challenge the ability of the UHS to provide sufficient inventory for 30
days. However, if the SR is not met, a plant shutdown is required.
Date of publication of individual notice in Federal Register:
September 5, 2014 (79 FR 53085).
Expiration date of individual notice: October 6, 2014 (public
comments); November 4, 2014 (hearing requests).
IV. Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
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.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: April 23, 2013, as supplemented by
letters dated June 19, and October 13, 2014.
Brief description of amendment: The amendment revised the Fermi 2
technical specification (TS) surveillance requirements (SRs) associated
with SR 3.8.4.2 and SR 3.8.4.5 to add a battery resistance limit; SR
3.8.6.3 to change the average electrolyte temperature of representative
cells, and SR 3.8.4.8 to change the frequency of battery capacity
testing.
Date of issuance: March 16, 2015.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 199. A publicly-available version is in ADAMS under
Accession No. ML15057A297; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-43: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 22, 2014 (79 FR
42542). The supplemental letters dated June 19, and October 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 a Safety Evaluation dated March 16, 2015.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: June 13, 2013, as supplemented by
letters dated August 28 and November 3, 2014, and January 22, 2015.
Brief description of amendment: The amendment revised the Technical
Specifications to risk-inform requirements regarding selected Required
Action end states by adopting Technical Specification Task Force
(TSTF)-423, Revision 1, ``Technical Specifications End States, NEDC-
32998-A,'' with some deviations as approved by the NRC staff. This
technical specification improvement is part of the Consolidated Line
Item Improvement Process (CLIIP). In addition, it approves a change to
the facility operating license for the River Bend Station, Unit 1. The
change deletes two license conditions that are no longer applicable and
adds a new license condition for maintaining commitments required for
the approval of this TSTF into the Updated Safety Analysis Report.
Date of issuance: February 17, 2015.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 185. A publicly-available version is in ADAMS under
Accession No. ML14106A167; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 20, 2013 (78 FR
51226). The supplemental letters dated August 28, and November 3, 2014,
and January 22, 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 17, 2015.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit 3, Westchester County, New York
Date of amendment request: February 4, 2014, as supplemented by
letter dated December 9, 2014.
Brief description of amendment: The amendment revised Technical
Specification 5.5.15, ``Containment Leakage Rate Testing Program,'' to
allow a permanent extension of the Type A primary containment
integrated leak
[[Page 17107]]
rate test frequency from once every 10 years to once every 15 years.
Date of issuance: March 13, 2015.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 256. A publicly-available version is in ADAMS under
Accession No. ML15028A308; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. DPR-64: The amendment revised the
Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2014 (79 FR
38587). The supplemental letter provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 13, 2015.
No significant hazards consideration comments received: No
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit 3, Westchester County, New York
Date of amendment request: April 1, 2014.
Brief description of amendment: The amendment revised Technical
Specification Figures 3.4.3-1, ``Heatup Limitations for Reactor Coolant
System,'' 3.4.3-2, ``Cooldown Limitations for Reactor Coolant System,''
and 3.4.3-3, ``Hydrostatic and Inservice Leak Testing Limitations for
Reactor Coolant System'' to address vacuum fill operations in the TSs.
The proposed changes clarify that the figures are applicable for vacuum
fill conditions where pressure limits are considered to be met for
pressures that are below 0 pounds per square inch gauge (psig) (i.e.,
up to and including full vacuum conditions). Vacuum fill operations for
the RCS can result in system pressures below 0 psig.
Date of issuance: March 6, 2015.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 255. A publicly-available version is in ADAMS under
Accession No. ML15050A144; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. DPR-64: The amendment revised the
Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: October 28, 2014 (79 FR
64223).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 6, 2015.
No significant hazards consideration comments received: No
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: April 5, 2013, as supplemented by letter
dated March 20, 2014.
Brief description of amendment: This amendment revised Technical
Specification (TS) 2.1.1 and 2.1.2, ``Safety Limits,'' by reducing the
reactor steam dome pressure from 785 pounds per square inch gauge
(psig) to 685 psig to resolve the Pressure Regulator Failure-Open
transient.
Date of issuance: March 12, 2015.
Effective date: As of the date of issuance, and shall be
implemented within 60 days of issuance.
Amendment No.: 242. A publicly-available version is in ADAMS under
Accession No. ML14272A070; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-35: Amendment revised
the License and TS.
Date of initial notice in Federal Register: August 6, 2013 (78 FR
47788). The supplement dated March 20, 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 the amendment is contained
in a Safety Evaluation dated March 12, 2015.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point
Nuclear Station, Unit 1, Oswego County, New York
Date of application for amendment: March 8, 2013, as supplemented
by letter dated May 16, 2013, July 8, July16, August 29, 2014, and
January 22, 2015. The public versions of these documents are available
in ADAMS at the Accession Nos. ML13073A103, ML13144A068, ML14203A050,
ML14199A384, ML14251A233, and ML15026A132, respectively.
Brief description of amendment: The amendment to the Nine Mile
Point Unit 1 (NMP1) Renewed Facility Operating License DPR-63 modified
Technical Specification (TS) Table 3.6.2i, ``Diesel Generator
Initiation,'' by revising the existing 4.16kV Power Board (PB) 102/103
Emergency Bus Undervoltage (Degraded Voltage) Operating Time value and
by updating the Set Point heading title. The TS revisions are being
made to resolve the green non-cited violation (NCV) associated with the
vital bus degraded voltage protection time delay documented in NRC
Inspection Report (IR) 05000220/201101, ``Nine Mile Point Nuclear
Station--NRC Unresolved Item Follow-up Inspection Report,'' dated
January 23, 2012 (ADAMS Accession No. ML12023A119), specifically,
NCV05000220/20 11011-01, ``Vital Bus Degraded Voltage Time Delay Not
Maintained within LOCA Analysis Assumptions.''
Date of issuance: March 12, 2015.
Effective date: effective as of the date of its issuance and shall
be implemented within 60 days.
Amendment No.: 217.
Renewed Facility Operating License No. DPR-63: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: June 11, 2013, (78 FR
35062). The supplements dated May 16, 2013, July 8, July16, August 29,
2014, and January 22, 2015, 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 initial
proposed no significant hazards consideration determination noticed in
the Federal Register on June 11, 2013 (78 FR 35062).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 2015.
No significant hazards consideration comments received: No
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: July 11, 2014, as supplemented
by letter dated December 1, 2014.
Brief description of amendments: The amendments incorporate several
administrative changes to the Facility Operating Licenses (FOLs) and
the Technical Specifications (TSs) such as deleting historical items
that are no longer applicable, correcting errors, and removing
references that are no longer valid.
Date of issuance: March 11, 2015.
[[Page 17108]]
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 296 and 299. A publicly-available version is in
ADAMS under Accession No. ML14363A227; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the FOLs and the TSs.
Date of initial notice in Federal Register: September 2, 2014 (79
FR 52062). The supplemental letter dated December 1, 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 March 11, 2015.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: October 18, 2013, as supplemented by
letters dated June 26, 2014, September 21, 2014, and February 4, 2015.
Brief description of amendments: The amendment changes the Beaver
Valley Power Station Technical Specifications (TS). Specifically, this
change request involves the adoption of an approved change to the
standard TS for Westinghouse plants (NUREG-1431), to allow relocation
of specific TS surveillance frequencies to a licensee-controlled
program. The proposed change is described in TS Task Force (TSTF)
Traveler, TSTF-425, Revision 3, ``Relocation Surveillance Frequencies
to Licensee Control--RITSTF [Risk-Informed Technical Specifications
Task Force] Initiative 5b'' (Agencywide Documents Access and Management
System (ADAMS) Accession No. ML090850642). A Notice of Availability was
published in the Federal Register on July 6, 2009 (74 FR 31996).
The proposed change relocates surveillance frequencies to a
licensee-controlled program, the Surveillance Frequency Control
Program. This change is applicable to licensees using probabilistic
risk guidelines contained in NRC-approved NEI 04-10, Revision 1,
``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed
Method for Control of Surveillance Frequencies'' (ADAMS Accession No.
ML071360456).
Date of issuance: March 6, 2015.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 292 and 179. A publicly-available version is in
ADAMS under Accession No. ML14322A461; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-66 and NPF-73:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2014 (79 FR
3416). The supplemental letters dated June 26, 2014, September 21,
2014, and February 4, 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 March 6, 2015.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: November 21, 2013, and supplemented by
the letters dated March 5 and June 30, 2014.
Brief description of amendment: The amendment authorizes changes to
the VEGP Units 3 and 4 Updated Final Safety Analysis Report to revise
the details of the effective thermal conductivity resulting from the
oxidation of the inorganic zinc component of the containment vessel
coating system.
Date of issuance: February 26, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 31. A publicly-available version is in ADAMS under
Accession No. ML15028A358; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: March 18, 2014 (79 FR
15150).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 26, 2015.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of application for amendment: September 25, 2012; as
supplemented on December 20, 2012; September 16, October 30, and
November 12, 2013; April 23, May 23, July 3, August 11, August 29, and
October 13, 2014; and January 16, 2015.
Brief description of amendments: The amendment authorizes the
transition of the Joseph M. Farley Nuclear Plant, Units 1 and 2, fire
protection program to a risk-informed, performance-based program based
on 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: March 10, 2015.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1-196, Unit 2-192. A publicly-available
version is in ADAMS under Accession No. ML14308A048, documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-2 and NPF-8: The amendments
revised the Renewed Facility Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: March 12, 2013 (78 FR
15750). The supplemental letters dated September 16, October 30, and
November 12, 2013; April 23, May 23, July 3, August 11, August 29, and
October 13, 2014; and January 16, 2015, provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 10, 2015.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 23rd day of March 2015.
[[Page 17109]]
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2015-07192 Filed 3-30-15; 8:45 am]
BILLING CODE 7590-01-P