[Federal Register Volume 80, Number 22 (Tuesday, February 3, 2015)]
[Notices]
[Pages 5798-5816]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-01917]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2015-0015]
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 January 8, 2015, to January 21, 2015. The
last biweekly notice was published on January 20, 2015.
DATES: Comments must be filed by March 5, 2015. A request for a hearing
must be filed by April 6, 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-2015-0015. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected].
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: 3WFN-06-A44M, 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: Beverly A. Clayton, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-3475, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2015-0015 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-2015-0015.
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-2015-0015 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, 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
[[Page 5799]]
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 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 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-
[[Page 5800]]
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.
Energy Northwest, Docket No. 50-397, Columbia Generating Station
(Columbia), Benton County, Washington
Date of amendment request: November 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14336A100.
Description of amendment request: The proposed amendment would
modify the Technical Specifications to revise values for the safety
limit minimum critical power ratio (SLMCPR) due to core loading fuel
management changes for the upcoming Columbia operating cycle.
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 basis of the Safety Limit Minimum Critical Power Ratio
(SLMCPR) is to ensure no mechanistic fuel damage is calculated to
occur if the limit is not violated. The new SLMCPR values preserve
the existing margin to transition boiling. The derivation of the
revised SLMCPR for Columbia, for incorporation into the Technical
Specifications and its use to determine plant and cycle-specific
thermal limits, has been performed using NRC approved methods. The
revised SLMCPR values do not change the method of operating the
plant and have no effect on the probability of an accident
initiating event or transient.
Based on the above, Energy Northwest has concluded that the
proposed change will not result in 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 analyzed?
Response: No.
The proposed changes result only from a specific analysis for
the Columbia core reload design. These changes do not involve any
new or different methods for operating the facility. No new
initiating events or transients result from these changes.
Based on the above, Energy Northwest has concluded that the
proposed change will not create the possibility of a new or
different kind of accident from those previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The new SLMCPR is calculated using NRC approved methods with
plant and cycle specific parameters for the current core design. The
SLMCPR value remains conservative enough to ensure that at least
99.9% of all fuel rods in the core will avoid transition boiling if
the limit is not violated,
[[Page 5801]]
thereby preserving the fuel cladding integrity. The operating limit
minimum critical power ratio (MCPR) is established to ensure that no
fuel damage results during anticipated operational occurrences
(AOOs). Accordingly, the margin of safety is maintained with the
revised values.
As a result, Energy Northwest has determined that the proposed
change will not result in a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
Acting NRC Branch Chief: Eric R. Oesterle.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
Date of amendment request: December 5, 2014. A publicly-available
version is in ADAMS under Accession No. ML14351A074.
Description of amendment request: The amendment would revise
Technical Specifications (TSs) Section 3.6.2.1, regarding containment
spray and cooling systems, by eliminating second completion times
limiting time from discovery of failure to meet a limiting condition
for operation (LCO). The proposed revision is consistent with NRC-
approved Technical Specifications Task Force (TSTF) Traveler TSTF-439,
Revision 2, ``Eliminate Second Completion Times Limiting Time from
Discovery of Failure to Meet an LCO'' (Adams Accession No.
ML051860296).
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 any accident previously
evaluated?
Response: No.
The proposed change that incorporated TSTF-439, Revision 2,
[will eliminate] certain Completion Times from the TS. 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 revised
Completion Times are no different [from] the consequences of the
same accident during the existing Completion Times. As a result, the
consequences of an accident previously evaluated are not affected by
this change. The proposed change does not alter or prevent the
ability of structures, systems, or components (SSCs) from performing
their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits.
The proposed change does not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed change does not increase the types or amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed change is consistent with the
[previous] safety analysis assumptions and resultant consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an 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 change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
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. The proposed change does not alter any 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 change involve a significant reduction in
the margin of safety?
Response: No.
The proposed change to delete the second Completion Times 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 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.
The NRC staff has reviewed the licensee's analysis and determined
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Branch Chief: Shana R. Helton.
Omaha Public Power District (OPPD), Docket No. 50-285, Fort Calhoun
Station, Unit 1, Washington County, Nebraska
Date of amendment request: December 26, 2014. A publicly-available
version is in ADAMS under Accession No. ML14365A123.
Description of amendment request: The proposed amendment upgrades
the Emergency Action Level (EAL) scheme by adopting NRC-endorsed
Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Methodology for the
Development of Emergency Action Levels for Non-Passive Reactors,''
issued January 2011 (ADAMS Accession No. ML110240324).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to OPPD's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, ``Development of
Emergency Action Levels for Non-Passive Reactors,'' do not reduce
the capability to meet the emergency planning requirements
established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed
changes do not reduce the functionality, performance, or capability
of OPPD's ERO [emergency response organization] to respond in
mitigating the consequences of any design basis accident.
The probability of a reactor accident requiring implementation
of Emergency Plan EALs has no relevance in determining whether the
proposed changes to the EALs reduce the effectiveness of the
Emergency Plans. As discussed in Section D, ``Planning Basis,'' of
NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of
Radiological Emergency Response Plans and Preparedness in Support of
Nuclear Power Plants'' [issued November 1980; ADAMS Accession No.
ML040420012]:
. . . The overall objective of emergency response plans is to
provide dose savings (and in some cases immediate life saving) for a
spectrum of accidents that could produce offsite doses in excess of
Protective Action Guides (PAGs). No single specific accident
sequence should be isolated as the one for which to plan because
each accident could have different consequences, both in nature and
degree. Further, the range of possible selection for a planning
basis is very large, starting with a zero point of requiring no
planning at all because significant offsite radiological accident
consequences are unlikely to occur, to planning for the worst
possible accident, regardless of its extremely low likelihood . . .
Therefore, OPPD did not consider the risk insights regarding any
specific accident
[[Page 5802]]
initiation or progression in evaluating the proposed changes.
The proposed changes do not involve any physical changes to
plant equipment or systems, nor do they alter the assumptions of any
accident analyses. The proposed changes do not adversely affect
accident initiators or precursors nor do they alter the design
assumptions, conditions, and configuration or the manner in which
the plant is operated and maintained. The proposed changes do not
adversely affect the ability of Structures, Systems, or Components
(SSCs) to perform their intended safety functions in mitigating 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 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 to OPPD's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any
physical changes to plant systems or equipment. The proposed changes
do not involve the addition of any new plant equipment. The proposed
changes will not alter the design configuration, or method of
operation of plant to be performed as required. The proposed changes
do not create any new credible failure mechanisms, malfunctions, or
accident initiators.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from those that have been
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to OPPD's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not alter or exceed a
design basis or safety limit. There is no change being made to
safety analysis assumptions, safety limits, or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed change. There are no changes to setpoints or
environmental conditions of any SSC or the manner in which any SSC
is operated. Margins of safety are unaffected by the proposed
changes to adopt the NEI 99-01, Revision 6, EAL scheme guidance. The
applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E
will continue to be met.
Therefore, the proposed changes do not involve any 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 A. Repka, Esq., Winston & Strawn, 1700
K Street NW., Washington, DC 20006-3817.
Acting NRC Branch Chief: Eric R. Oesterle.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028,
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: July 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14202A088.
Description of amendment request: The proposed changes would revise
the Combined Licenses (COLs) by (1) providing additional detail to
describe the mechanical connection between the internal containment
structural module steel faceplates and the base concrete, (2) allowing
for increases in the thickness of the structural wall module
faceplates, (3) identifying changes to the wall thicknesses for
portions of some internal containment structural wall modules, and (4)
identifying the use of steel plates, structural shapes, reinforcement
bars, or tie bars between the faceplates of the structural wall
modules, where needed to meet applicable code requirements.
Because this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 Design Control
Document (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 design function of the internal containment structures is to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in those structures.
These structures are structurally designed to meet seismic Category
I requirements as defined in Regulatory Guide 1.29.
The changes to the design details for the structural modules do
not have an adverse impact on the response of the nuclear island
structures to safe shutdown earthquake ground motions or loads due
to anticipated transients or postulated accident conditions, nor do
they change the seismic Category I classification. Evaluations have
been performed which determined that the proposed changes do not
have a significant impact on the calculated loads for the affected
structural modules, or critical locations, and no significant impact
on the global seismic model. The changes to the design details for
the structural modules do not impact the support, design, or
operation of mechanical and fluid systems. There is no change to
plant systems or the response of systems to postulated accident
conditions. There is no change to the predicted radioactive releases
due to postulated accident conditions. The plant response to
previously evaluated accidents or external events is not adversely
affected, nor does the change described create any new accident
precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are to revise design details for the
internal containment structural modules. The changes do not change
the design requirements of the nuclear island structures, nor do
they change the seismic Category I classification. The changes to
the design details for the internal containment structural modules
do not change the design function, support, design, or operation of
mechanical and fluid systems. The changes to the design details for
the internal containment structural modules do not result in a new
failure mechanism for the nuclear island structures or introduce any
new accident precursors. As a result, the design function of the
nuclear island structures is not adversely affected by the proposed
change.
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 requested amendment proposes changes to the structural
details associated with the in-containment structural modules. The
purpose of these changes is to ensure that the requirements
contained in the applicable construction codes are met. As discussed
in UFSAR [Updated Final Analysis Report], Section 3.8.3.5, ``Design
Procedures and Acceptance Criteria,'' the in-containment structural
modules are designed in accordance with ACI [American Concrete
Institute] 349 and AISC [American Institute of Steel Construction]
N690. Thus, the identification of additional structural module
connection details, the increase in structural module faceplate and
wall thicknesses, and the addition of additional reinforcement in
specific areas are proposed to ensure that the codes of record, and
the associated margins contained therein, continue to be met as
specified in the design basis. Structural and seismic analysis of
the modified sections in accordance with the methodologies
identified in the UFSAR has confirmed that the applicable
requirements of ACI 349 and AISC N690 continue to be met for
affected in-containment structural modules.
As a result, the proposed changes do not adversely affect any
safety related equipment
[[Page 5803]]
or other design functions, design code compliance, 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.
Therefore, the requested 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 J. Burkhart.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of amendment request: December 19, 2014. A publicly-available
version is in ADAMS Package Accession No. ML14363A422.
Description of amendment request: The licensee proposes to expand
the emergency planning zone (EPZ) boundary, to revise the evacuation
time estimates (ETA) analysis, and revise the alert and notification
system (ANS) design reports to encompass the expanded EPZ boundary.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes, which include expansion of the EPZ
boundary and revision of the ETE analysis and ANS design reports to
encompass the expanded EPZ boundary, do not impact the physical
function of plant structures, systems, or components (SSC) or the
manner in which SSCs perform their design function. The proposed
changes neither adversely affect accident initiators or precursors,
nor alter design assumptions. The proposed changes do not alter or
prevent the ability of SSCs to perform their intended function to
mitigate the consequences of an initiating event within assumed
acceptance limits. No operating procedures or administrative
controls that function to prevent or mitigate accidents are affected
by the proposed changes. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be installed
or removed) or a change in the method of plant operation. The
proposed changes will not introduce failure modes that could result
in a new accident, and the change does not alter assumptions made in
the safety analysis. The proposed changes, which include expansion
of the EPZ boundary and revision of the ETE analysis and ANS design
reports to encompass the expanded EPZ boundary, are not initiators
of any accidents. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is associated with the ability of the fission
product barriers (i.e., fuel cladding, reactor coolant system
pressure boundary, and containment structure) to limit the level of
radiation dose to the public. The proposed changes, which include
expansion of the EPZ boundary and revision of the ETE analysis and
ANS design reports to encompass the expanded EPZ boundary, do not
impact operation of the plant or its response to transients or
accidents. The proposed changes do not alter requirements of the
Technical Specifications or the Unit 1 Operating License. The
proposed changes do not involve a change in the method of plant
operation and no accident analyses will be affected by the proposed
changes.
Additionally, the proposed changes will 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 these proposed changes. The proposed changes will not result in
plant operation in a configuration outside the design basis. The
proposed changes do not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina
Electric & Gas Company, Post Office Box 764, Columbia, SC 29218.
NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: January 8, 2015. A publicly-available
version is in ADAMS under Accession No. ML15008A466.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4 by
departing from the plant-specific Design Control Document (DCD) Tier 1
(and corresponding Combined License Appendix C information) and Tier 2
material by making changes to specify the use of latching control
relays in lieu of breakers to de-energize the control rod drive
mechanism (CRDM) motor generator (MG) set generator field on a diverse
actuation system (DAS) signal.
Because this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 DCD, the licensee
also requested an exemption from the requirements of the Generic DCD
Tier 1 in accordance with 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 change to use field control relays in lieu of field
circuit breakers to de-energize the CRDM MG Set excitation field
does not result in a change to the basic MG Set design function,
which is to supply reliable electrical power to the CRDMs while
providing a trip function on a DAS signal, allowing the control rods
to drop. The Probabilistic Risk Assessment (PRA) is not adversely
affected. No safety-related structure, system, or component (SSC) or
function is adversely affected. The change does not involve nor
interface with any SSC accident initiator or initiating sequence of
events, and thus, the probabilities of the accidents evaluated in
the UFSAR are not affected. Because the change maintains the CRDM MG
set trip function used to mitigate an accident, the consequences of
the accidents evaluated in the UFSAR are not affected.
Therefore, there is no significant increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
[[Page 5804]]
There is no safety-related SSC or function adversely affected by
this proposed change to use control relays instead of breakers to
de-energize the CRDM MG set generator field on demand. This proposed
change does not change any equipment qualification or fission
product barrier. The change does not result in a new failure mode,
malfunction or sequence of events that could affect safety or
safety-related equipment. This activity will not allow for a new
fission product release path, result in a new fission product
barrier failure mode, or create a new sequence of events that would
result in significant fuel cladding failures.
Therefore, this activity does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
There is no safety-related SSC or function adversely affected by
this proposed change to use relays instead of breakers to control
the CRDM MG set generator field. The function to trip the MG set
generator field on a DAS signal, allowing the control rods to drop,
is not adversely affected by the use of relays as the device to de-
energize the generator field. The proposed change does not affect
any safety-related design code, function, design analysis, safety
analysis input or result, or design/safety margin. No safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by the requested change, thus, no margin of safety is
reduced.
Therefore, the proposed 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: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: November 24, 2014. A publicly-available
version is in ADAMS under Accession Package No. ML14335A689.
Description of amendment request: The licensee requested 24
revisions to the Technical Specifications. Twenty two revisions adopt
various previously NRC approved Technical Specifications Task Force
Travelers and two revisions are not associated with Travelers. A list
of the requested revisions is included in Enclosure 1 of the
application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration for each of the 24 changes requested, which is presented
below:
Request No. 1: TSTF-27-A, Revision 3, ``Revise SR Frequency for
Minimumn Temperature for Criticality''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Surveillance Frequency for
monitoring RCS temperature to ensure the minimum temperature for
criticality is met. The Frequency is changed from a 30 minute
Frequency when certain conditions are met to a periodic Frequency
that it is controlled in accordance with the Surveillance Frequency
Control Program. The measurement of RCS [reactor coolant system]
temperature is not an initiator of any accident previously
evaluated. The minimum RCS temperature for criticality is not
changed. As a result, the mitigation of any accident previously
evaluated is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises the Surveillance Frequency for
monitoring RCS temperature to ensure the minimum temperature for
criticality is met. The current, condition based Frequency
represents a distraction to the control room operator during the
critical period of plant startup. RCS temperature is closely
monitored by the operator during the approach to criticality and
temperature is recorded on charts and computer logs. Allowing the
operator to monitor temperature as needed by the situation and
logging RCS temperature at a periodic Frequency that it is
controlled in accordance with the Surveillance Frequency Control
Program is sufficient to ensure that the LCO [limiting condition for
operation] is met while eliminating a diversion of the operator's
attention.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 2: TSTF-46-A, Revision 1, ``Clarify the CIV Surveillance to
Apply Only to Automatic Isolation Valves''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the requirements in Technical
Specification SR 3.6.3.4, and the associated Bases, to delete the
reference to verifying the isolation time of ``each power operated''
containment isolation valve (CIV) and only require verification of
each ``automatic power operated containment isolation valve.'' The
closure times for CIVs that do not receive an automatic closure
signal are not an initiator of any design basis accident or event,
and therefore the proposed change does not increase the probability
of any accident previously evaluated. The CIVs are used to respond
to accidents previously evaluated. Power operated CIVs that do not
receive an automatic closure signal are not assumed to close in a
specified time. The proposed change does not change how the plant
would mitigate an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a change in the manner in
which the CIVs provide plant protection or introduce any new or
different operational conditions. Periodic verification that the
closure times for CIVs that receive an automatic closure signal are
within the limits established by the accident analysis will continue
to be performed under SR 3.6.3.4. The change does not alter
assumptions made in the safety analysis, and is consistent with the
safety analysis assumptions and current plant operating practice.
There are also no design changes associated with the proposed
changes, and the change does not involve a physical alteration of
the plant (i.e., no new or different type of equipment will be
installed).
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides clarification that only CIVs that
receive an automatic
[[Page 5805]]
isolation signal are within the scope of the SR 3.6.3.4. The
proposed change does not result in a change in the manner in which
the CIVs provide plant protection. Periodic verification that
closure times for CIVs that receive an automatic isolation signal
are within the limits established by the accident analysis will
continue to be performed. The proposed change does not affect the
safety analysis acceptance criteria for any analyzed event, nor is
there a change to any Safety Analysis Limit. The proposed change
does not alter the manner in which safety limits, limiting safety
system settings or limiting conditions for operation are determined,
nor is there any adverse effect on those plant systems necessary to
assure the accomplishment of protection functions. The proposed
change will not result in plant operation in a configuration outside
the design basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 3: TSTF-87-A, Revision 2, ``Revise ``RTBs Open'' and ``CRDM
Deenergized'' Actions to ``Incapable of Rod Withdrawal''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change revises the Required Actions for LCO 3.4.5, ``RCS
Loops--Mode 3,'' Conditions C.2 and D.1, from ``De-energize all
control rod drive mechanisms,'' to ``Place the Rod Control System in
a condition incapable of rod withdrawal.'' It also revises LCO
3.4.9, ``Pressurizer,'' Required Action A. 1, from requiring the
Reactor Trip Breakers to be open after reaching MODE 3 to ``Place
the Rod Control System in a condition incapable of rod withdrawal,''
and to require full insertion of all rods. Inadvertent rod
withdrawal can be an initiator for design basis accidents or events
during certain plant conditions, and therefore must be prevented
under those conditions. The proposed Required Actions for LCO 3.4.5
and LCO 3.4.9 satisfy the same intent as the current Required
Actions, which is to prevent inadvertent rod withdrawal when an
applicable Condition is not met, and is consistent with the
assumptions of the accident analysis. As a result, the proposed
change does not increase the probability of any accident previously
evaluated. The proposed change does not change how the plant would
mitigate an accident previously evaluated as in both the current and
proposed requirements, rod withdrawal is prohibited.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change provides less specific, but equivalent,
direction on the manner in which inadvertent control rod withdrawal
is to be prevented when the Conditions of LCO 3.4.5 and LCO 3.4.9
are not met. Rod withdrawal will continue to be prevented when the
applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. There are
no design changes associated with the proposed changes, and the
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed). The change
does not alter assumptions made in the safety analysis, and is
consistent with the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides the operational flexibility of
allowing alternate, but equivalent, methods of preventing rod
withdrawal when LCO 3.4.5 and LCO 3.4.9 are not met. The proposed
change does not affect the safety analysis acceptance criteria for
any analyzed event, nor is there a change to any safety analysis
limit. The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined, nor is there any adverse effect on those
plant systems necessary to assure the accomplishment of protection
functions. The proposed change will not result in plant operation in
a configuration outside the design basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 4: TSTF-245-A, Revision 1, ``AFW Train Operable When in
Service''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the requirements in Technical
Specification 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to
clarify the operability of an AFW train when it is aligned for
manual steam generator level control. The AFW System is not an
initiator of any design basis accident or event, and therefore the
proposed change does not increase the probability of any accident
previously evaluated. The AFW System is used to respond to accidents
previously evaluated. The proposed change does not affect the design
of the AFW System, and no physical changes are made to the plant.
The proposed change does not significantly change how the plant
would mitigate an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a change in the manner in
which the AFW System provides plant protection. The AFW System will
continue to supply water to the steam generators to remove decay
heat and other residual heat by delivering at least the minimum
required flow rate to the steam generators. There are no design
changes associated with the proposed changes, and the change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed). The change does not
alter assumptions made in the safety analysis, and is consistent
with the safety analysis assumptions and current plant operating
practice. Manual control of AFW level control valves is not an
accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change provides the operational flexibility of
allowing an AFW train(s) to be considered operable when it is not in
the normal standby alignment and is temporarily incapable of
automatic initiation, such as during alignment and operation for
manual steam generator level control, provided it is capable of
being manually realigned to the AFW heat removal mode of operation.
The proposed change does not result in a change in the manner in
which the AFW System provides plant protection. The AFW System will
continue to supply water to the steam generators to remove decay
heat and other residual heat by delivering at least the minimum
required flow rate to the steam generators. The proposed change does
not affect the safety analysis acceptance criteria for any analyzed
event, nor is there a change to any Safety Analysis Limit. The
proposed change does not alter the manner in which safety limits,
limiting safety system settings 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.
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
[[Page 5806]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Request No. 5: TSTF-247-A, Revision 0, ``Provide Separate Condition
Entry for Each PORV and Block Valve''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the requirements in Technical
Specification 3.4.11, ``Pressurizer PORVs [power operated relief
valves],'' to clarify that separate Condition entry is allowed for
each block valve. Additionally, the Actions are modified to no
longer require that the PORVs be placed in manual operation when
both block valves are inoperable and cannot be restored to operable
status within the specified Completion Time. This preserves the
overpressure protection capabilities of the PORVs. The pressurizer
block valves are used to isolate their respective PORV in the event
it is experiencing excessive leakage, and are not an initiator of
any design basis accident or event. Therefore the proposed change
does not increase the probability of any accident previously
evaluated. The PORV and block valves are used to respond to
accidents previously evaluated. The proposed change does not affect
the design of the PORV and block valves, and no physical changes are
made to the plant. The proposed change does not change how the plant
would mitigate an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not result in a change in the manner in
which the PORV and block valves provide plant protection. The PORVs
will continue to provide overpressure protection, and the block
valves will continue to provide isolation capability in the event a
PORV is experiencing excessive leakage. There are no design changes
associated with the proposed changes, and the change does not
involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed). The change does not
alter assumptions made in the safety analysis, and is consistent
with the safety analysis assumptions and current plant operating
practice. Operation of the PORV block valves is not an accident
initiator.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes provide clarification that separate
Condition entry is allowed for each block valve. Additionally, the
Actions are modified to no longer require that the PORVs be placed
in manual operation when both block valves are inoperable and cannot
be restored to operable status within the specified Completion Time.
This preserves the overpressure protection capabilities of the
PORVs. The proposed change does not result in a change in the manner
in which the PORV and block valves provide plant protection. The
PORVs will continue to provide overpressure protection, and the
block valves will continue to provide isolation capability in the
event a PORV is experiencing excessive leakage. The proposed change
does not affect the safety analysis acceptance criteria for any
analyzed event, nor is there a change to any safety analysis limit.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined, nor is there any adverse effect on those
plant systems necessary to assure the accomplishment of protection
functions. The proposed change will not result in plant operation in
a configuration outside the design basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 6: TSTF-248-A, Revision 0, ``Revise Shutdown Margin
Definition for Stuck Rod Exception''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the definition of Shutdown Margin
to eliminate the requirement to assume the highest worth control rod
is fully withdrawn when calculating Shutdown Margin if it can be
verified by two independent means that all control rods are
inserted. The method for calculating shutdown margin is not an
initiator of any accident previously evaluated. If it can be
verified by two independent means that all control rods are
inserted, the calculated Shutdown Margin, without the conservatism
of assuming the highest worth control rod is withdrawn, is accurate
and consistent with the assumptions in the accident analysis. As a
result, the mitigation of any accident previously evaluated is not
affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change modifies the definition of Shutdown Margin
to eliminate the requirement to assume the highest worth control rod
is fully withdrawn when calculating Shutdown Margin if it can be
verified by two independent means that all control rods are
inserted. The additional margin of safety provided by the assumption
that the highest worth control rod is fully withdrawn is unnecessary
if it can be independently verified that all controls rods are
inserted.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 7: TSTF-266-A, Revision 3, ``Eliminate the Remote Shutdown
System Table of Instrumentation and Controls''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change removes the list of Remote Shutdown System
instrumentation and controls from the Technical Specifications and
places them in the Bases. The Technical Specifications continue to
require that the instrumentation and controls be operable. The
location of the list of Remote Shutdown System instrumentation and
controls is not an initiator to any accident previously evaluated.
The proposed change will have no effect on the mitigation of any
accident previously evaluated because the instrumentation and
controls continue to be required to be operable.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
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
[[Page 5807]]
kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change removes the list of Remote Shutdown System
instrumentation and controls from the Technical Specifications and
places it in the Bases. The review performed by the NRC when the
list of Remote Shutdown System instrumentation and controls is
revised will no longer be needed unless the criteria in 10 CFR 50.59
are not met such that prior NRC review is required. The Technical
Specification requirement that the Remote Shutdown System be
operable, the definition of operability, the requirements of 10 CFR
50.59, and the Technical Specifications Bases Control Program are
sufficient to ensure that revision of the list without prior NRC
review and approval does not introduce a significant safety risk.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 8: TSTF-272-A, Revision 1, ``Refueling Boron Concentration
Clarification''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the Applicability of Specification
3.9.1, ``Boron Concentration,'' to clarify that the boron
concentration limits are only applicable to the refueling canal and
the refueling cavity when those volumes are attached to the Reactor
Coolant System (RCS). The boron concentration of water volumes not
connected to the RCS are not an initiator of an accident previously
evaluated. The ability to mitigate any accident previously evaluated
is not affected by the boron concentration of water volumes not
connected to the RCS.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change modifies the Applicability of Specification
3.9.1, ``Boron Concentration,'' to clarify that the boron
concentration limits are only applicable to the refueling canal and
the refueling cavity when those volumes are attached to the RCS.
Technical Specification SR 3.0.4 requires that Surveillances be met
prior to entering the Applicability of a Specification. As a result,
the boron concentration of the refueling cavity or the refueling
canal must be verified to satisfy the LCO prior to connecting those
volumes to the RCS. The margin of safety provided by the refueling
boron concentration is not affected by this change as the RCS boron
concentration will continue to satisfy the LCO.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 9: TSTF-273-A, Revision 2, ``Safety Function Determination
Program Clarifications''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed TS changes add explanatory text to the programmatic
description of the Safety Function Determination Program (SFDP) in
Specification 5.5.15 to clarify in the requirements that
consideration does not have to be made for a loss of power in
determining loss of function. The Bases for LCO 3.0.6 is revised to
provide clarification of the ``appropriate LCO for loss of
function,'' and that consideration does not have to be made for a
loss of power in determining loss of function. The changes are
editorial and administrative in nature, and therefore do not
increase the probability of any accident previously evaluated. No
physical or operational changes are made to the plant. The proposed
change does not change how the plant would mitigate an accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are editorial and administrative in nature
and do not result in a change in the manner in which the plant
operates. The loss of function of any specific component will
continue to be addressed in its specific TS LCO and plant
configuration will be governed by the required actions of those
LCOs. The proposed changes are clarifications that do not degrade
the availability or capability of safety related equipment, and
therefore do not create the possibility of a new or different kind
of accident from any accident previously evaluated. There are no
design changes associated with the proposed changes, and the changes
do not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed). The changes do not
alter assumptions made in the safety analysis, and are consistent
with the safety analysis assumptions and current plant operating
practice. Due to the administrative nature of the changes, they
cannot be an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to TS 5.5.15 are clarifications and are
editorial and administrative in nature. No changes are made the LCOs
for plant equipment, the time required for the TS Required Actions
to be completed, or the out of service time for the components
involved. The proposed changes do not affect the safety analysis
acceptance criteria for any analyzed event, nor is there a change to
any safety analysis limit. The proposed changes do not alter the
manner in which safety limits, limiting safety system settings or
limiting conditions for operation are determined, nor is there any
adverse effect on those plant systems necessary to assure the
accomplishment of protection functions. The proposed changes will
not result in plant operation in a configuration outside the design
basis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
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.
Request No. 10: TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode
Restriction Notes''
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 four
diesel generator (DG) Surveillances to allow performance of the
Surveillance in whole or in part to reestablish DG Operability. The
emergency diesel generators 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
reestablish
[[Page 5808]]
Operability of a DG. The consequences of an accident previously
evaluated during the period that the DG is being tested to
reestablish Operability are no different from the consequences of an
accident previously evaluated while the 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 it's assumed safety function. The proposed
change will only allow the performance of the Surveillances to
reestablish Operability and the proposed changes may not be used to
remove a DG from service. In addition, the proposed change will
potentially shorten the time that a DG is unavailable because
testing to reestablish Operability can be performed without a plant
shutdown. 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.
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.
Request No. 11: TSTF-284-A, Revision 3, ``Add `Met vs. Perform' to
Technical Specification 14, Frequency''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes insert a discussion paragraph into
Specification 1.4, and several new examples are added to facilitate
the use and application of SR Notes that utilize the terms ``met''
and ``perform''. The changes also modify SRs in multiple
Specifications to appropriately use ``met'' and ``perform''
exceptions. The changes are administrative in nature because they
provide clarification and correction of existing expectations, and
therefore the proposed change does not increase the probability of
any accident previously evaluated. No physical or operational
changes are made to the plant. The proposed change does not
significantly change how the plant would mitigate an accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are administrative in nature and do not
result in a change in the manner in which the plant operates. The
proposed changes provide clarification and correction of existing
expectations that do not degrade the availability or capability of
safety related equipment, and therefore do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. There are no design changes associated with
the proposed changes, and the changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed). The changes do not alter assumptions made in the
safety analysis, and are consistent with the safety analysis
assumptions and current plant operating practice. Due to the
administrative nature of the changes, they cannot be an accident
initiator.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes are administrative in nature and do not
result in a change in the manner in which the plant operates. The
proposed changes provide clarification and correction of existing
expectations that do not degrade the availability or capability of
safety related equipment, or alter their operation. The proposed
changes do not affect the safety analysis acceptance criteria for
any analyzed event, nor is there a change to any safety analysis
limit. The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined, nor is there any adverse effect on those
plant systems necessary to assure the accomplishment of protection
functions. The proposed changes will not result in plant operation
in a configuration outside the design basis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
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.
Request No. 12: TSTF-308-A, Revision 1, ``Determination of Cumulative
and Projected Dose Contributions in RECP''
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.
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.
Request No. 13: TSTF-312-A, Revision 1, ``Administrative Control of
Containment Penetrations''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would allow containment penetrations to be
unisolated under administrative controls during core
[[Page 5809]]
alterations or movement of irradiated fuel assemblies within
containment. The status of containment penetration flow paths (i.e.,
open or closed) is not an initiator for any design basis accident or
event, and therefore the proposed change does not increase the
probability of any accident previously evaluated. The proposed
change does not affect the design of the primary containment, or
alter plant operating practices such that the probability of an
accident previously evaluated would be significantly increased. The
proposed change does not significantly change how the plant would
mitigate an accident previously evaluated, and is bounded by the
fuel handling accident (FHA) analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Allowing penetration flow paths to be open is not an initiator
for any accident. The proposed change to allow open penetration flow
paths will not affect plant safety functions or plant operating
practices such that a new or different accident could be created.
There are no design changes associated with the proposed changes,
and the change does not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed). The
change does not alter assumptions made in the safety analysis, and
is consistent with the safety analysis assumptions and current plant
operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
TS 3.9.3 provides measures to ensure that the dose consequences
of a postulated FHA inside containment are minimized. The proposed
change to LCO 3.9.3 will allow penetration flow path(s) to be open
during refueling operations under administrative control. These
administrative controls will provide assurance that prompt closure
of open penetrations flow paths can and will be achieved in the
event of an FHA inside containment, and will minimize dose
consequences. The proposed change is bounded by the existing FHA
analysis. The proposed change does not affect the safety analysis
acceptance criteria for any analyzed event, nor is there a change to
any safety analysis limit. The proposed change does not alter the
manner in which safety limits, limiting safety system settings or
limiting conditions for operation are determined, nor is there any
adverse effect on those plant systems necessary to assure the
accomplishment of protection functions. The proposed change will not
result in plant operation in a configuration outside the design
basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 14: TSTF-314-A, Revision 0, ``Require Static and Transient
FQ Measurement''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Required Actions of
Specification 3.1.4, ``Rod Group Alignment Limits,'' and
Specification 3.2.4, ``Quadrant Power Tilt Ratio,'' to require
measurement of both the steady state and transient portions of the
Heat Flux Hot Channel Factor, FQ(Z). This change will
ensure that the hot channel factors are within their limits when the
rod alignment limits or quadrant power tilt ratio are not within
their limits. The verification of hot channel factors is not an
initiator of any accident previously evaluated. The verification
that both the steady state and transient portion of FQ(Z)
are within their limits will ensure this initial assumption of the
accident analysis is met should a previously evaluated accident
occur.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises the Required Actions in the
Specifications for Rod Group Alignment Limits and Quadrant Power
Tilt Ratio to require measurement of both the steady state and
transient portions of the Heat Flux Hot Channel Factor,
FQ(Z). This change is a correction that ensures that the
plant conditions are as assumed in the accident analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 15: TSTF-315-A, Revision 0, ``Reduce Plant Trips Due to
Spurious Signals to the NIS During Physics Testing''
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 3.1.8, ``PHYSICS TESTS
Exceptions--MODE 2,'' to allow the number of channels required by
LCO 3.3.1, ``RTS Instrumentation,'' to be reduced from ``4'' to
``3'' to allow one nuclear instrumentation channel to be used as an
input to the reactivity computer for physics testing without placing
the nuclear instrumentation channel in a tripped condition. A
reduction in the number of required nuclear instrumentation channels
is not an initiator to any accident previously evaluated. With the
nuclear instrumentation channel placed in bypass instead of in trip,
reactor protection is provided by the intermediate range neutron
flux detectors and the nuclear instrumentation system operating in a
two-out-of-three channel logic. As a result, the ability to mitigate
any accident previously evaluated is not significantly affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change reduces the probability of a spurious
reactor trip during physics testing. The reactor trip system
continues to be capable of protecting the reactor utilizing the
intermediate range neutron flux reactor trip and the power range
neutron flux trips operating in a two-out-of-three trip logic. As a
result, the reactor is protected and the probability of a spurious
reactor trip is significantly reduced.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 5810]]
amendment request involves no significant hazards consideration.
Request No. 16: TSTF-325, Revision 0, ``ECCS Conditions and Required
Actions with Less Than 100% Equivalent ECCS Flow''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change corrects the structure of Technical
Specification 3.5.2 to assure its proper application. There is no
change in intent or in the way the Technical Specification is
applied. The literal (and unintended) interpretation of the existing
LCO structure could, under some circumstances, provide longer than
intended Completion Times for restoration of operability. The
proposed change only clarifies the requirements of the Required
Actions. Since the proposed change affects neither the Technical
Specification intent, nor its application, the proposed change will
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 corrects the structure of the Technical
Specification to assure its correct application. There is no change
in intent or in the way the Technical Specification is applied. The
proposed changes would not result in any physical alterations to the
plant configuration, no new equipment is added, no equipment
interfaces are modified, and no changes to any equipment's function
or the method of operating the equipment are being made. As the
proposed changes would not change the design, configuration or
operation of the plant, no new or different kinds of accident modes
are created.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change corrects the structure of the Technical
Specification to assure its correct application. There is no change
in intent or in the way the Technical Specification is applied.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Request No. 17: TSTF-340-A, Revision 3, ``Allow 7 Day Completion Time
for a Turbine-Driven AFW Pump Inoperable''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Specification 3.7.5, ``Auxiliary
Feedwater (AFW) System,'' to allow a 7 day Completion Time to
restore an inoperable turbine-driven pump in Mode 3 immediately
following a refueling outage, if Mode 2 has not been entered. An
inoperable AFW turbine-driven pump is not an initiator of any
accident previously evaluated. The ability of the plant to mitigate
an accident is no different while in the extended Completion Time
than during the existing Completion Time.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises Specification 3.7.5, ``Auxiliary
Feedwater (AFW) System,'' to allow a 7 day Completion Time to
restore an inoperable turbine-driven AFW pump in Mode 3 immediately
following a refueling outage if Mode 2 has not been entered. In Mode
3 immediately following a refueling outage, core decay heat is low
and the need for AFW is also diminished. The two operable motor
driven AFW pumps are available and there are alternate means of
decay heat removal if needed. As a result, the risk presented by the
extended Completion Time is minimal.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 18: TSTF-343, Revision 1, ``Containment Structural
Integrity''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise the Technical Specifications (TS)
Administrative Controls programs for consistency with the
requirements of 10 CFR 50, paragraph 55a(g)(4) for components
classified as Code Class CC. The proposed changes affect the
frequency of visual examinations that will be performed for the
concrete surfaces of the containment for the purpose of the
Containment Leakage Rate Testing Program, and allows those
examinations to be performed during power operation in addition to
during a refueling outage.
The frequency of visual examinations of the containment and the
mode of operation during which those examinations are performed does
not affect the initiation of any accident previously evaluated. The
use of NRC approved methods and frequencies for performing the
inspections will ensure the containment continues to perform the
mitigating function assumed for accidents previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes revise the TS Administrative Controls
programs for consistency with the requirements of 10 CFR 50,
paragraph 55a(g)(4) for components classified as Code Class CC. The
proposed changes affect the frequency of visual examinations that
will be performed for the concrete surfaces of the containment for
the purpose of the Containment Leakage Rate Testing Program, and
allows those examinations to be performed during power operation in
addition to during a refueling outage.
The proposed changes do not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed changes will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes revise the Technical Specifications (TS)
Administrative Controls programs for consistency with the
requirements of 10 CFR 50, paragraph 55a(g)(4) for components
classified as Code Class CC. The proposed changes affect the
frequency of visual examinations that will be performed for the
concrete surfaces of the containment for the purpose of the
Containment Leakage Rate Testing Program,
[[Page 5811]]
and allows those examinations to be performed during power operation
in addition to during a refueling outage. The safety function of the
containment as a fission product barrier will be maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 19: TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5
Allowing Shutdown Cooling Loops Removal from Operation''
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 an LCO Note to LCO 3.9.5, ``RHR and
Coolant Circulation--Low Water Level,'' to allow securing the
operating train of Residual Heat Removal (RHR) for up to 15 minutes
to support switching operating trains. The allowance is restricted
to conditions in which core outlet temperature is maintained at
least 10 degrees F below the saturation temperature, when there are
no draining operations, and when operations that could reduce the
reactor coolant system (RCS) boron concentration are prohibited.
Securing an RHR train to facilitate the changing of the operating
train is not an initiator to any accident previously evaluated. The
restrictions on the use of the allowance ensure that an RHR train
will not be needed during the 15 minute period to mitigate any
accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change adds an LCO Note to LCO 3.9.5, ``RHR and
Coolant Circulation--Low Water Level,'' to allow securing the
operating train of RHR to support switching operating trains. The
allowance is restricted to conditions in which core outlet
temperature is maintained at least 10 degrees F below the saturation
temperature, when there are no draining operations, and when
operations that could reduce the reactor coolant system (RCS) boron
concentration are prohibited. With these restrictions, combined with
the short time frame allowed to swap operating RHR trains and the
ability to start an operating RHR train if needed, the occurrence of
an event that would require immediate operation of an RHR train is
extremely remote.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Request No. 20: TSTF-355-A, Revision 0, ``Changes to RTS and ESF
Tables''
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The RTS [Reactor Trip System] and ESFAS [Engineered Safety
Feature Actuations System] instrument functions are part of the
accident mitigation response and are not themselves an initiator of
any accident previously evaluated. Therefore, the probability of an
accident previously evaluated is not significantly affected by the
proposed changes. The changes ensure that automatic protective
actions will be initiated at or before the condition assumed in the
safety analysis, and are in accordance with the intent of the
Technical Specifications. The proposed changes will not cause any
design or analysis acceptance criteria to be exceeded. Since there
will be no adverse effect on the trip setpoints or the
instrumentation associated with the trip setpoints, there will be no
significant increase in the 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 include modifications to the format of the
nominal trip setpoints that preserve safety analysis assumptions
related to accident mitigation. The protection system will continue
to initiate the protective actions as assumed in the safety
analysis. The proposed changes will continue to ensure that the trip
setpoints are maintained consistent with the setpoint methodology
and the plant safety analysis. As the proposed changes do not change
the design, configuration or operation of the plant, no new or
different kinds of accident modes are created.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not alter any nominal trip setpoints,
allowable values, or limiting safety system settings, and will
continue to ensure that the trip setpoints are maintained consistent
with the setpoint methodology and the plant safety analysis. The
response of protection systems to accident transients reported in
the Final Safety Analysis Report is unaffected by this change, and
accident analysis acceptance criteria are consequently not affected.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Request No. 21: TSTF-371-A, Revision 1, ``NIS Power Range Channel Daily
SR TS Change to Address Low Power Decalibration''
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 3.3.1, ``RTS
Instrumentation,'' Surveillances 3.3.1.2 and 3.3.1.3 to move
requirements currently in a Note to the Surveillance itself. The
change in presentation is editorial and does not affect the
application of the Surveillances. The proposed change does not
affect 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, 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 Specification 3.3.1, ``RTS
Instrumentation,'' Surveillances 3.3.1.2 and 3.3.1.3 to move
requirements currently in a Note to the Surveillance itself. The
proposed change represents an editorial preference and does not
affect the
[[Page 5812]]
performance of the Surveillance or plant operation. The safety
function tested by the Surveillance is unaffected.
Therefore, this 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.
Request No. 22: TSTF-439-A, Revision 2, ``Eliminate Second Completion
Times Limiting Time From Discovery of Failure To Meet an LCO''
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 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.
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.
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.
Request No. 23: ISTS Adoption #1--Revise LCO 3.3.2 ESFAS Interlock P-4
Required Action Completion Time
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 Condition to be entered when the
ESFAS Interlock P-4 is inoperable. Current Technical Specifications
require restoring the channel to Operable status within 24 hours or
be in Mode 3 within the next 12 hours and Mode 5 within the
following 52 hours. The proposed change provides 48 hours to restore
the inoperable channel, or be in Mode 3 in 54 hours and Mode 4 in 60
hours. The ESFAS P-4 interlock is not an initiator to any accident
previously evaluated. The consequences of any accident previously
evaluated during the proposed Completion Time are no different from
the consequences during the existing Completion Time. As a result,
the proposed change does not result in a significant increase in the
consequences of any accident previously evaluated.
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 provides an additional 24 hours to restore
an inoperable ESFAS P-4 Interlock. During the proposed Completion
Time, manual actions can perform the functions provided by the
inoperable P-4 interlock. Also, the proposed Completion Time is
reasonable given the available redundant channel, and the low
probability of an event occurring during this interval.
Therefore, this 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.
Request No. 24: Revise LCO 3.5.5 to 8-hour Completion Time and Note
allowance
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the LCO 3.5.5, ``Seal Injection
Flow,'' Action A, ``Seal injection flow not within limit,''
Completion Time from 4 hours to 8 hours and the Note to SR 3.5.5.1
to allow 8 hours instead of 4 hours to stabilize reactor coolant
system (RCS) pressure prior to verifying the seal injection throttle
valves are properly adjusted. The proposed change does not involve
the addition or removal of any equipment, or any design changes to
the facility. Seal injection flow is not an initiator of any
accident previously evaluated. The consequences of any accident
previously evaluated during the extended Completion Time or Note
allowance are the same as during the existing Completion Time and
Note allowance.
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 provides additional time to verify seal
injection flow is within limit or to restore seal injection flow to
within limit if it is discovered that it is not within limit. The
additional time is acceptable on the basis that there is little
likelihood of an event that would challenge the ECCS occurring
during the 8-hour window, and it reduces the pressure on the
operations staff should iterations in the adjustment procedure be
necessary to balance seal injection flow.
Therefore, this 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: Leigh D. Perry, SVP & General Counsel of
Operations and Nuclear, Southern Nuclear
[[Page 5813]]
Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J. Pascarelli.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: October 2, 2014. A publicly-available
version is in ADAMS under Accession No. ML14275A441.
Description of amendment request: The proposed amendment upgrades
the Emergency Action Level scheme by adopting NRC-endorsed Nuclear
Energy Institute 99-01, Revision 6, ``Methodology for the Development
of Emergency Action Levels for Non-Passive Reactors,'' issued January
2011 (ADAMS Accession No. ML110240324).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the Callaway Plant emergency action
levels do not impact the physical function of plant structures,
systems, or components (SSC) or the manner in which SSCs perform
their design function. The proposed changes neither adversely affect
accident initiators or precursors, nor alter design assumptions. The
proposed changes do not alter or prevent the ability of SSCs to
perform their intended function to mitigate the consequences of an
initiating event within assumed acceptance limits. No operating
procedures or administrative controls that function to prevent or
mitigate accidents are affected by the proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be installed
or removed) or a change in the method of plant operation. The
proposed changes will not introduce failure modes that could result
in a new accident, and the change does not alter assumptions made in
the safety analysis. The proposed changes to the Callaway Plant
emergency action levels are not initiators of any accidents.
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.
Margin of safety is associated with the ability of the fission
product barriers (i.e., fuel cladding, reactor coolant system
pressure boundary, and containment structure) to limit the level of
radiation dose to the public. The proposed changes do not impact
operation of the plant or its response to transients or accidents.
The changes do not affect the Technical Specifications or the
operating license. The proposed changes do not involve a change in
the method of plant operation, and no accident analyses will be
affected by the proposed changes. Additionally, the proposed changes
will 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 these changes. The proposed changes
will not result in plant operation in a configuration outside the
design basis. The proposed changes do not adversely affect systems
that respond to safely shut down the plant and to maintain the plant
in a safe shutdown condition. The emergency plan will continue to
activate an emergency response commensurate with the extent of
degradation of plant 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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
Acting NRC Branch Chief: Eric R. Oesterle.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating 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.
Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: April 24, 2014.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3/4.4.5, ``Steam Generator Tube Integrity,'' TS
6.8.4.I, ``Steam Generator Program,'' and TS 6.9.1.7, ``Steam Generator
Tube Inspection Report'' to address implementation associated with the
inspections and reporting requirements as described in Technical
Specifications Task Force (TSTF) TSTF-510, Revision 2, ``Revision to
Steam Generator Program Inspection Frequencies and Tube Sample
Selection.''
Date of issuance: January 9, 2015.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 145. A publicly-available version is in ADAMS under
Accession No. ML14307A800; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-63 The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 22, 2014 (79 FR
42543).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 9, 2015.
No significant hazards consideration comments received: No.
[[Page 5814]]
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: October 31, 2013, as
supplemented by letters dated May 29, 2014, and September 9, 2014.
Brief description of amendment: The amendment revised Technical
Specification Surveillance Requirements 3.5.1.4 and 3.5.2.5 for low
pressure core spray and low pressure coolant injection pump flows.
Date of issuance: January 7, 2015.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 229. A publicly-available version is in ADAMS under
Accession No. ML14335A189; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 8, 2014 (79 FR
19399). The supplemental letters dated May 29, 2014, and September 9,
2014, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 7, 2015.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2, Pope County, Arkansas
Date of application for amendment: January 21, 2014, as
supplemented by letters dated March 17 and September 24, 2014.
Brief description of amendment: The amendment revised the Technical
Specification 6.5.16 requirements for the local leak test required for
the containment building emergency escape air lock doors, in that it
would require a seal contact verification in lieu of the current seal
pressure test to verify leak tightness.
Date of issuance: January 22, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 299. A publicly-available version is in ADAMS under
Accession No. ML14350B285; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: April 15, 2014 (79 FR
21296). The supplemental letter dated September 24, 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 January 22, 2015.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: November 15, 2013, as supplemented by
letters dated April 16, 2014; September 11, 2014; and November 7, 2014.
Brief description of amendments: The amendments revise the
Technical Specification (TS) requirements related to the response time
for the main steam line flow-high isolation function.
Date of issuance: January 7, 2015.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 214 and 175. A publicly-available version is in
ADAMS under Accession No. ML14344A681; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-39 and NPF-85:
Amendments revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: February 4, 2014 (79 FR
6642). The supplemental letters dated April 16, 2014; September 11,
2014; and November 7, 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 January 7, 2015.
No significant hazards consideration comments received: No.
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 requests: July 16, 2013, as supplemented by
letters dated September 18, 2013, January 22, April 7, August 12, and
November 11, 2014.
Brief description of amendments: The amendments revises the
Technical Specifications to include the use of neutron absorbing spent
fuel pool rack inserts (i.e., NETCO-SNAP-IN[supreg] rack inserts) for
the purpose of criticality control in the spent fuel pools.
Date of issuance: December 31, 2014.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 253-Unit 1; 248-Unit 2. A publicly-available
version is in ADAMS under Accession No. ML14346A306; documents related
to these amendments are listed in the safety evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-29 and DPR-30: The
amendments revised the Technical Specifications and Facility Operating
License.
Date of initial notice in Federal Register: July 8, 2014 (79 FR
38577). The supplemental letters dated September 18, 2013, January 22,
April 7, August 12, and November 11, 2014, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 31, 2014.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: November 14, 2013.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.5.11, ``Primary Containment Leakage Rate Testing
Program,'' by removing TS 5.5.11.d.2.b, the reduced pressure testing
option for drywell airlock door leakage testing. This testing
methodology is not required and does not reflect the current testing
practice at MNGP. As such, the drywell
[[Page 5815]]
airlock door seals will be tested by performing an overall airlock
leakage test as specified in current TS 5.5.11.d.2.a.
Date of issuance: January 8, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 187. A publicly-available version is in ADAMS under
Accession No. ML14323A033; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-22: This amendment
revises the Renewed Facility Operating License and the Technical
Specifications.
Date of initial notice in Federal Register: August 5, 2014 (79 FR
45478).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 8, 2015.
No significant hazards consideration comments received: No.
South Carolina Electric and Gas Company Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: April 3, 2014, as supplemented by letter
dated May 19, 2014.
Brief description of amendment: The amendment revises Tier 2*
information, incorporated into the VCSNS Units 2 and 3 Updated Final
Safety Analysis Report (UFSAR). Specifically, the amendment revises the
details regarding the structural floor of the Auxiliary Building and
its constructability. Notes are added to drawings in Subsection 3H.5 of
the UFSAR in order to clarify variations in detail design such as size
and spacing or reinforcement and spans of the noncritical sections of
floors.
Date of issuance: July 18, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 14. A publicly-available version is in ADAMS under
Accession No. ML14188B185; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: April 29, 2014 (79 FR
24024).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 18, 2014.
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: March 17, 2014, and revised by letters
dated May 8, September 2, and October 2, 2014.
Brief description of amendment: The amendment revises the VEGP
Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) by
clarifying how human diversity was applied during the design process
for the Component Interface Module and Diverse Actuation System. The
changes to the VEGP Units 3 and 4 UFSAR include changes to Table 1.6,
``Material Referenced,'' Chapter 7, Sections 7.1.2.14.1, 7.1.7 and
7.2.4 and the addition of Appendix 7A to Chapter 7. The changes to the
VEGP Units 3 and 4 UFSAR modify information related to human diversity,
as presented in a Tier 2* document, WCAP-17179-P and WCAP-17179-NP,
``AP1000 Component Interface Module Technical Report,'' Revision 2, and
two Tier 2 documents, WCAP-15775, ``AP1000 Instrumentation and Control
Defense-in-Depth and Diversity Report,'' Revision 4 and WCAP-17184-P,
``AP1000 Diverse Actuation System Planning and Functional Design
Summary Technical Report,'' that are incorporated by reference in the
VEGP Units 3 and 4 UFSAR.
Date of issuance: December 24, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 28. A publicly-available version is in ADAMS under
Accession No. ML14329A298; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: April 29, 2014 (79 FR
24021).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 24, 2014.
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: August 22, 2014, and revised by letter
dated September 23, 2014, and supplemented by letters dated October 30
and November 6, 2014.
Brief description of amendment: The amendment revises the VEGP
Units 3 and 4 Updated Final Safety Analysis Report to reflect changes
related to:
(a) Installation of an additional non-safety-related battery;
(b) Revision to the annex building internal configuration by
converting a shift turnover room to a battery room, adding an
additional battery equipment room, and moving a fire area wall;
(c) Increase in the height of a room in the annex building; and
(d) Increase in thicknesses of certain annex building floor slabs.
In addition, the proposed changes also include reconfiguring
existing rooms and related rooms, wall, and access path changes and
making changes to the corresponding Tier 1 information in Appendix C to
the Combined Licenses.
Date of issuance: December 23, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 27. A publicly-available version is in ADAMS under
Accession No. ML14323A609; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: October 14, 2014 (79 FR
61662).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 2014.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: January 23, 2014.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.4.12, ``Cold Overpressure Mitigation System
(COMS),'' to reflect the mass input transient analysis that assumes an
Emergency Core Cooling System centrifugal charging pump and the normal
charging pump capable of injecting into the reactor coolant system when
TS 3.4.12 is applicable. The amendment also revised TS Table 3.3.1-1,
``Reactor Trip System Instrumentation,'' to remove unnecessary page
number references.
Date of issuance: January 20, 2015.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 210. A publicly-available version is in ADAMS under
[[Page 5816]]
Accession No. ML14350B239; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 1, 2014 (79 FR
18348).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 20, 2015.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 26th day of January 2015.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2015-01917 Filed 2-2-15; 8:45 am]
BILLING CODE 7590-01-P