[Federal Register Volume 78, Number 63 (Tuesday, April 2, 2013)]
[Notices]
[Pages 19746-19757]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-07467]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0060]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
Background
Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission publish notice of any amendments issued, or proposed to be
issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 7, 2013, to March 20, 2013. The last
biweekly notice was published on March 19, 2013 (78 FR 16876).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and is publicly available, by
searching on http://www.regulations.gov under Docket ID NRC-2013-0060.
You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0060. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0060 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly available, by the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0060.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2013-0060 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 that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC
[[Page 19747]]
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.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in section 50.92 of Title 10 of the Code
of Federal Regulations (10 CFR), this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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 regulations are accessible electronically from the NRC Library on
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC'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 10
days prior to the filing deadline, the participant should contact the
Office of
[[Page 19748]]
the Secretary by email at [email protected], or by telephone at
301-415-1677, to request (1) a digital identification (ID) certificate,
which allows the participant (or its counsel or representative) to
digitally sign documents and access the E-Submittal server for any
proceeding in which it is participating; and (2) advise the Secretary
that the participant will be submitting a request or petition for
hearing (even in instances in which the participant, or its counsel or
representative, already holds an NRC-issued digital ID certificate).
Based upon this information, the Secretary will establish an electronic
docket for the hearing in this proceeding if the Secretary has not
already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the 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 agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[email protected], or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
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 following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected].
[[Page 19749]]
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit 2, New London County, Connecticut
Date of amendment request: July 21, 2010.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) \3/4\.9.3.1, ``Decay Time'' for
Millstone Power Station Unit 2 (MPS2). The proposed change would revise
TS \3/4\.9.3.1 by reducing the minimum decay time for irradiated fuel
prior to movement in the reactor vessel from 150 hours to 100 hours. A
reduction in the minimum decay time requirement is requested to provide
additional flexibility in outage planning such that irradiated fuel can
be moved from the reactor vessel to the spent fuel pool earlier in an
outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
Will operation of the facility in accordance with the proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The accident of concern related to the proposed change is the
FHA [fuel handling accident]. This accident assumes a dropped fuel
assembly with resulting damage and release of the gap activity from
the entire assembly. The FHA assumes that fuel movement is delayed
for some time period after shutdown to accommodate for radioactive
decay of the short-lived fission products. The probability of a FHA
occurrence is dependent on moving fuel not when the fuel movement
occurs. Reducing the decay time required by TS \3/4\.9.3.1 from 150
hours to 100 hours does not increase the probability of a FHA since
the timing of fuel movement in the reactor pressure vessel does not
alter/impact the manner in which fuel assemblies are handled.
Reducing the decay time requirement in TS \3/4\.9.3.1 from 150
hours to 100 hours does not change the consequences of the offsite
dose and control room dose projections for the currently approved
design basis FHA analysis. The current FHA analysis presented in
FSAR [final safety analysis report] Section 14.7.4 and approved in
License Amendment 298 assumes a minimum 100 hour decay time.
Therefore, the dose results of this FHA analysis are unchanged, and
remain within applicable regulatory limits.
Based on the reasons presented above, operation of the facility
in accordance with the proposed amendment would not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2
Will operation of the facility in accordance with the proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated?
Response: No.
The proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
No new accident will be created as a result of reducing the decay
time requirement in TS \3/4\.9.3.1. Plant operation, including fuel
handling, will not be affected by the proposed change, as to when
fuel is moved and no new failure modes will be created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
Criterion 3
Will operation of the facility in accordance with the proposed
change involve a significant reduction in the margin of safety?
Response: No.
The proposed change does not significantly reduce the margin of
safety. The current analysis of record for the FHA already accounts
for irradiated fuel with at least 100 hours of decay. This approved
analysis has shown that the projected doses will remain within
applicable regulatory limits; therefore, the margin of safety is
unchanged.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Acting Branch Chief: Sean C. Meighan.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating, Unit 2, Westchester County, New York
Date of amendment request: January 28, 2013.
Description of amendment request: Nuclear Safety Advisory Letter
11-5 identified Westinghouse methodology errors in the long-term mass
and energy releases during a large break loss-of-coolant accident.
These impacted the containment integrity analysis for Indian Point,
Unit 2. A re-analysis of the large break loss-of-coolant accident for
the limiting single failure concluded that four, rather than three
containment fan cooler units would need to be credited. The proposed
change will revise Technical Specification Bases Sections 3.6.4,
``Containment Pressure,'' 3.6.5, ``Containment Air Temperature,'' and
3.6.6, ``Containment Spray System and Containment Fan Cooler Unit (FCU)
System.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously
identified?
Response: No.
The proposed change would not change the current limiting EDG
[emergency diesel generator] failure but would credit four rather
than three can cooler units for containment heat removal. Four fan
cooler units are available after the single failure. The fan cooler
units are not accident initiators so the probability of an accident
does not increase. Crediting all four fan cooler units will keep the
post accident containment pressure within current limits and
therefore does not increase the probability or consequences of a
previously evaluated accident, but is a change from the analyses
approved by the NRC [Nuclear Regulatory Commission] during stretch
power uprate.
Therefore the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new of different
kind of accident from any accident previously evaluated?
Response: No.
There are no changes to design, no changes to operating
procedures, and the revised licensing basis change is consistent
with the available equipment following the postulated worst case
single failure.
Therefore the proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The change reflects the credit for equipment that was always
available but not previously credited (as a conservatism) in the
licensing basis analyses. With credit for four fan cooler units, the
post accident containment pressure remains within current limits and
there is no reduction in a margin of safety.
Therefore the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440
[[Page 19750]]
Hamilton Avenue, White Plains, NY 10601.
NRC Acting Branch Chief: Sean Meighan.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating, Unit 2, Westchester County, New York
Date of amendment request: February 6, 2013.
Description of amendment request: The proposed amendment will
revise the Reactor Heatup and Cooldown curves and Low Temperature
Overpressure Protection Requirements in Technical Specifications (TSs)
3.4.3, ``RCS [reactor coolant system] Pressure and Temperature (P/T)
Limits,'' 3.4.6, ``RCS Loops--MODE 4,'' 3.4.7, ``RCS Loops--MODE 5,
Loops Filled,'' 3.4.10, ``Pressurizer Safety Valves,'' and 3.4.12,
``Low Temperature Overpressure Protection (LTOP).''
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. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence of consequences of an accident previously
evaluated.
The proposed TS changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Except for a setpoint change for automatic PORV [power-operated
relief valve] actuation, there are no physical changes to the plant
being introduced by the proposed changes to the heatup and cooldown
limitation curves. The proposed changes do not modify the RCS
pressure boundary. That is, there are no changes in operating
pressure, materials, or seismic loading. The proposed changes do not
adversely affect the integrity of the RCS pressure boundary such
that its function in the control of radiological consequences is
affected. The proposed heatup and cooldown limitation curves were
generated in accordance with the fracture toughness requirements of
10CFR50 [10 CFR 50] Appendix G, and ASME B&PV code [American Society
of Mechanical Engineers Boiler and Pressure Vessel Code], Section
XI, Appendix G edition with 2000 Addenda. The proposed heatup and
cooldown limitation curves were established in compliance with the
methodology used to calculate and predict effects of radiation on
embrittlement of RPV [reactor pressure vessel] beltline materials.
Use of this methodology provides compliance with the intent of
10CFR50 [10 CFR 50] Appendix G and provides margins of safety that
ensure non-ductile failure of the RPV will not occur. The proposed
heatup and cooldown limitation curves prohibit operation in regions
where it is possible for non-ductile failure of carbon and low alloy
RCS materials to occur. Hence, the primary coolant pressure boundary
integrity will be maintained throughout the limit of applicability
of the curves, 48 EFPY [Effective Full Power Years].
Operation within the proposed LTOP limits ensures that
overpressurization of the RCS at low temperatures will not result in
component stresses in excess of those allowed by the ASME B&PV Code
Section XI Appendix G.
Consequently, the proposed changes do not involve a significant
increase in the probability or the consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No new modes of operation are introduced by the proposed
changes. The proposed changes will not create any failure mode not
bounded by previously evaluated accidents. Further, the proposed
changes to the heatup and cooldown limitation curves and the LTOP
limits do not affect any activities or equipment other than the RCS
pressure boundary and do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Consequently, the proposed changes do not create the possibility
of a new or different kind of accident, from any accident previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
The Proposed TS changes do not involve a significant reduction
in the margin of safety. The revised heatup and cooldown limitation
curves and LTOP limits are established in accordance with current
regulations and the ASME B&PV Code 1998 edition with 2000 Addenda.
These proposed changes are acceptable because the ASME B&PV Code
maintains the margin of safety required by 10CFR50.55(a) [10 CFR
50.55(a)]. Because operation will be within these limits, the RCS
materials will continue to behave in a non-brittle manner consistent
with the original design bases.
The proposed changes to the allowable operation of charging and
safety injection pumps when LTOP is required to be operable is
consistent with the IP2 licensing bases as established in TS
Amendment 262.
Therefore, Entergy has concluded that 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, and with the changes noted above in square brackets, 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. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: Sean Meighan.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating, Unit 3, Westchester County, New York
Date of amendment request: January 28, 2013.
Description of amendment request: Nuclear Safety Advisory Letter
(NSAL) 11-5 identified Westinghouse methodology errors in the long-term
mass and energy releases during a large break loss-of-coolant accident.
These impacted the containment integrity analysis for Indian Point Unit
No. 3 and required revisions to limiting initial operating conditions
(i.e., containment temperature, containment pressure, and refueling
water storage tank temperature) and require revisions to Technical
Specifications (TSs) 3.5.4, ``Refueling Water Storage Tank (RWST),''
and 3.6.4, ``Containment Pressure.'' In addition, revisions are
proposed for TS 3.6.3, ``Containment Isolation Valves,'' to delete a
redundant surveillance requirement and TS 5.5.15, ``Containment Leakage
Rate Testing Program,'' to reflect a slightly higher calculated
containment peak pressure.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would not change the current EDG [emergency
diesel generator] failure but limits the RWST temperature to
<=105[emsp14][deg]F and containment pressure to <=1.5 psig [pounds
per square inch gauge] (when RWST temperature is >95[emsp14][deg]F
or containment/accumulator temperature is >125[emsp14][deg]F). The
proposed change also removes a redundant TS for Containment testing
and corrects the peak pressure in the containment testing program.
The initial conditions assumed in accident analysis are not accident
initiators so the probability of an accident does not increase. The
change in initial conditions compensates for the error corrections
and maintains the post accident containment pressure within 0.38
psig of the current value and within Containment testing limits and
therefore does not increase the probability or consequences of a
previously evaluated accident. Therefore the proposed change does
not involve a significant increase in the
[[Page 19751]]
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The change to the initial conditions assumed in the analysis for
peak containment pressure, the removal of a redundant Technical
Specification and the correction to the peak pressure limit in the
Containment testing program do not create the possibility of a new
or different accident. There are no changes to design or operating
procedures that could create a new or different kind of accident
since the changes only affect the initiating conditions. The revised
analysis is consistent with the available equipment following the
postulate worst case single failure.
Therefore the proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The change in peak containment pressure is from 42 psig to 42.38
psig as a result of the error corrections of NSAL-11-5 and change to
the initial conditions for the RWST temperature and containment
pressure. There is an insignificant impact on other programs due to
change in peak containment pressure, which remains well below the
containment design pressure of 47 psig. Therefore there is not
significant reduction in margin.
Therefore the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: Sean Meighan.
Exelon Generation Company (EGC), LLC, Docket No. 50-374, LaSalle County
Station (LSCS), Unit 2, LaSalle County, Illinois
Date of amendment request: October 15, 2012.
Description of amendment request: The proposed amendment would
remove License Conditions which are no longer necessary to address an
interim configuration of the LaSalle County Station, Unit 2, spent fuel
pool prior to completed installation of NETCO-SNAP-IN[supreg] inserts.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change removes License Conditions within the LSCS
Unit 2 Operating License related to interim configurations of the
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and
the required completion date for installation. All changes proposed
by EGC in this license amendment request are administrative in
nature because they remove License Conditions that have either been
satisfied or that are no longer applicable. There are no physical
changes to the facilities, nor any changes to the station operating
procedures, limiting conditions for operation, or limiting safety
system settings.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change removes License Conditions within the LSCS
Unit 2 Operating License related to interim configurations of the
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and
the required completion date for installation. There are no changes
to the SFP criticality analysis associated with the proposed change.
No physical changes to the plant are proposed, and there are no
changes to the manner in which the plant is operated. Rather, the
proposed change is administrative because it involves removing
License Conditions that have either been satisfied or that are no
longer applicable.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change removes License Conditions within the LSCS
Unit 2 Operating License related to interim configurations of the
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and
the required completion date for installation. Plant safety margins
are established through limiting conditions for operation, limiting
safety system settings, and safety limits specified in Technical
Specifications. The proposed change does not alter these established
safety margins. The proposed change does not alter the criticality
analysis for the SFP and does not affect the SFP criticality safety
margin. The proposed change is administrative because it involves
removing License Conditions that have either been satisfied or that
are no longer applicable.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Tamra Domeyer, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Acting Branch Chief: Jeremy S. Bowen.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: July 12, 2012.
Description of amendment request: The proposed amendments would
modify Technical Specification 3.7.3, ``Ultimate Heat Sink,'' by
establishing controls which allow for the increase of cooling water
temperature from 104[emsp14][deg]F to 107[emsp14][deg]F for plant
safety systems.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change makes no physical changes to the plant, nor
does it alter any of the assumptions or conditions upon which the
UHS is designed. These assumptions and conditions as described in
the LSCS UFSAR include failure of the cooling lake dike, a loss of
offsite power, and a DBA LOCA on one unit and a normal shutdown of
the other unit.
The accidents analyzed in the UFSAR are assumed to be initiated
by the failure of plant structures, systems, or components (SSCs).
An inoperable UHS is not an initiator of any analyzed events as
described in the UFSAR. The impact on the structural integrity of
the UHS due to a potential increase water temperature prior to and
during the UHS design basis event has been evaluated, and does not
increase the probability of the failure of the cooling lake dike.
The proposed temperature limit for cooling water supplied to the
plant from the CSCS Pond could reduce the commercial capability of
the LSCS units; however, it does not result in an increase in the
probability of occurrence for any of the events described in the
UFSAR.
The basis provided in Regulatory Guide 1.27, ``Ultimate Heat
Sink for Nuclear Power Plants,'' Revision 1, dated March 1974, was
employed for the temperature analysis of the LSCS UHS to implement
General Design Criteria 2, ``Design bases for protection
[[Page 19752]]
against natural phenomena,'' and 44, ``Cooling water,'' of Appendix
A to 10 CFR Part 50. This Regulatory Guide was employed for both the
original design and licensing basis of the LSCS UHS and a subsequent
evaluation which investigated the potential for changing the average
water temperature of the cooling water supplied to the plant from
the CSCS Pond from a fixed temperature limit to a limit based on the
time of day. The meteorological conditions chosen for the LSCS UHS
analysis utilized a 31-day period consisting of the most severe one
day, combined with the most severe 30 days based on historical data.
The heat loads selected for the UHS analysis considered failure of
the cooling lake dike, a loss of offsite power, and a DBA LOCA on
one unit and a normal shutdown of the other unit. The LSCS cooling
lake is conservatively assumed to be unavailable at the start of the
event.
The analysis shows that with an initial UHS temperature less
than or equal to the proposed time-of-day-based limit, the required
safety-related heat loads can be adequately cooled for 30 days while
continuing to ensure safety-related cooling water temperature
remains less than the design temperature for LSCS, Units 1 and 2.
Based on the above, it has been demonstrated that the change of
the initial temperature limit for cooling water supplied to the
plant from the CSCS Pond to less than or equal to a temperature
based on the time of day will not impede the ability of the
equipment and components cooled by the UHS during a UHS design basis
event to perform their safety functions.
There is no impact of this change on LSCS safety analyses
including the consequences of all postulated events since all
required safety-related equipment continues to perform as designed.
The effects of the proposed change on the ability of the UHS to
assure that a 30-day supply of water is available considering losses
due to evaporation, seepage, and firefighting have been considered.
Sufficient inventory remains available to mitigate the design basis
event for the LSCS UHS for the required 30-day period.
Therefore, the proposed activity does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not physically alter the operation,
testing, or maintenance of any plant SSCs beyond operating with a
UHS temperature limit based on the time of day. The proposed change
is bounded by existing design analyses. Moreover, the UHS
temperature does not initiate accident precursors. The impact of
increased UHS temperature can affect the commercial operation of the
plant, but the proposed change would not create any accident not
considered in the LSCS UFSAR.
This proposed change will not alter the manner in which
equipment operation is initiated, nor will the functional demands on
credited equipment be changed. No alteration in the procedures that
ensure the LSCS units remain within analyzed limits is proposed, and
no change is being made to procedures relied upon to respond to an
off-normal event.
As such, no new failure modes are being introduced. The proposed
change does not alter assumptions made in the LSCS safety analysis.
Changing the temperature of cooling water supplied to the plant
from the CSCS Pond (i.e., the UHS) as proposed has no impact on
plant accident response. The proposed temperature limits do not
introduce new failure mechanisms for SSCs. An engineering analysis
performed to support the change in temperature of cooling water
supplied to the plant from the CSCS Pond provides the basis to
conclude that the equipment is adequately designed for operation as
proposed.
All systems that are important to safety will continue to be
operated and maintained within their design bases, and the proposed
change will continue to ensure that all associated systems and
components are operated reliably within their design capabilities.
The proposed change will ensure the maximum temperature of the
cooling water supplied to the plant during the UHS design basis
event remains less than the current safety-related cooling water
design temperature for LSCS, Units 1 and 2. Therefore, there is no
impact of this change on the LSCS safety analyses including
inventory and cooling requirements for safety-related systems using
the UHS as their cooling water supply.
All systems will continue to be operated within their design
capabilities, no new failure modes are introduced, nor is there any
adverse impact on plant equipment; therefore, the proposed change
does not result in the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is determined by the design and
qualification of the plant equipment, the operation of the plant
within analyzed limits, and the point at which protective or
mitigative actions are initiated. The proposed change does not
impact any of these factors. There are no required design changes or
equipment performance parameter changes associated with the proposed
change. No protection setpoints are affected as a result of this
change. The proposed change in the limit for the temperature of
cooling water supplied to the plant from the CSCS Pond will not
change the operational characteristics of the design of any
equipment or system. All accident analysis assumptions and
conditions will continue to be met.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Tamra Domeyer, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Acting Branch Chief: Jeremy S. Bowen.
NextEra Energy Seabrook, LLC., Docket No. 50-443, Seabrook Station,
Unit 1, Rockingham County, New Hampshire
Date of amendment request: March 1, 2013.
Description of amendment request: The proposed amendment will
revise the Seabrook Technical Specifications (TSs). The proposed
amendment will make administrative changes and corrections to the TSs.
The proposed changes delete TS Index and make corrections to TS 3.4.8,
``Reactor Coolant System Specific Activity,'' and TS 6.8.1.6.a, ``Core
Operating Limits Report.''
Basis for proposed NSHC 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes (1) remove the index from the TS, (2)
correct an error in the units of activity for 100/E in TS 3.4.8,
Reactor Coolant System Specific Activity, and (3) remove an
incorrect, non-applicable reference in TS 6.8, Core Operating Limits
Report. The proposed changes are all administrative in nature. The
administrative changes are not initiators of any accident previously
evaluated, and, consequently, the probability and consequences of an
accident previously evaluated is not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes are administrative in nature so no new or
different accidents result from the proposed changes. The changes do
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed), a significant change
in the method of plant operation, or new operator actions. The
changes do not alter 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. The proposed changes do not involve a significant reduction
in the margin of safety.
[[Page 19753]]
Margin of safety is associated with confidence in 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 administrative
changes do not involve a change in the method of plant operation, do
not affect any accident analyses, and do not relax any safety system
settings.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves NSHC.
Attorney for licensee: Mr. James Petro, Managing Attorney, Florida
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Meena Khanna.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2,
Goodhue County, Minnesota
Date of amendment request: September 28, 2012, as supplemented on
November 8, 2012 and December 18, 2012.
Description of amendment request: The proposed amendment requests
U.S. Nuclear Regulatory Commission (NRC) approval to adopt a new fire
protection licensing basis which complies with the requirements in 10
CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide
(RG) 1.205, Revision 1, ``Risk-Informed, Performance Based Fire
Protection for Existing Light-Water Nuclear Power Plants.'' This
amendment request also follows the guidance in Nuclear Energy Institute
(NEI) 04-02, Revision 2, ``Guidance for Implementing a Risk-Informed,
Performance-Based Fire Protection Program Under 10 CFR 50.48(c).'' If
approved, the PINGP fire protection program would transition to a new
Risk-Informed, Performance-Based alternative in accordance with 10 CFR
50.48(c), which incorporates by reference National Fire Protection
Association Standard 805 (NFPA 805). The NFPA 805 fire protection
program would supersede the current fire protection program licensing
basis in accordance with 10 CFR Part 50, Appendix R.
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.
Operation of the Prairie Island Nuclear Generating Plant (PINGP)
in accordance with the proposed amendment does not increase the
probability or consequences of accidents previously evaluated.
Engineering analyses, which may include engineering evaluations,
probabilistic safety assessments, and fire modeling evaluations,
have been performed to demonstrate that the performance-based
requirements of National Fire Protection Association Standard 805
(NFPA 805) have been satisfied. The PINGP Updated Safety Analysis
Report (USAR) documents the analyses of design basis accidents
(DBAs) at PINGP. The proposed amendment does not adversely affect
accident initiators nor alter design assumptions, conditions, or
configurations of the facility that would increase the probability
or consequences of accidents previously evaluated. Further, the
changes to be made for fire hazard protection and mitigation do not
adversely affect the ability of structures, systems, and components
(SSCs) to perform their design functions, nor do they affect the
postulated initiators or assumed failure modes for accidents
described and evaluated in the USAR. SSCs required to safely shut
down the reactor and to maintain it in a safe shutdown condition
will remain capable of performing their design functions.
The purpose of this proposed amendment is to permit PINGP to
adopt a new fire protection licensing basis which complies with the
requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision
1 of Regulatory Guide (RG) 1.205. The NRC considers that NFPA 805
provides an acceptable methodology and performance criteria for
licensees to identify fire protection systems and features that are
an acceptable alternative to the 10 CFR Part 50, Appendix R fire
protection features (69 FR 33536; June 16, 2004). Engineering
analyses, in accordance with NFPA 805, have been performed to
demonstrate that the risk-informed, performance-based (RI-PB)
requirements per NFPA 805 have been met.
NFPA 805, taken as a whole, provides an acceptable alternative
to 10 CFR 50.48(b), satisfies 10 CFR 50.48(a) and General Design
Criterion (GDC) 3 of Appendix A to 10 CFR Part 50, and meets the
underlying intent of the NRC's existing fire protection regulations
and guidance, and provides for defense-in-depth. The goals,
performance objectives, and performance criteria specified in
Chapter 1 of NFPA 805 ensure that if there are any increases in the
net core damage frequency (CDF) or risk associated with this license
amendment request (LAR) submittal, the increase will be small and
consistent with the Commission's Safety Goal Policy.
Based on this, the implementation of this amendment does not
significantly increase the probability of any accident previously
evaluated. Equipment required to mitigate an accident remains
capable of performing the assumed function(s). The proposed
amendment will not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of any accident previously evaluated.
Therefore, the consequences of any accident previously evaluated
are not significantly increased with the implementation of the
proposed amendment.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Operation of PINGP in accordance with the proposed amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated. Any scenario or
previously analyzed accident with offsite dose was included in the
evaluation of DBAs documented in the USAR. The proposed change does
not alter the requirements or function for systems required during
accident conditions. Implementation of the new fire protection
licensing basis which complies with the requirements in 10 CFR
50.48(a) and (c) and the guidance in Revision 1 of RG 1.205 will not
result in new or different accidents.
The proposed amendment does not introduce new or different
accident initiators nor alter design assumptions or conditions of
the facility. The proposed amendment does not adversely affect the
ability of SSCs to perform their design function. SSCs required to
safely shut down the reactor and maintain it in a safe shutdown
condition remain capable of performing their design functions.
The purpose of this amendment is to permit PINGP to adopt a new
fire protection licensing basis which complies with the requirements
in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG
1.205. The NRC considers that NFPA 805 provides an acceptable
methodology and performance criteria for licensees to identify fire
protection systems and features that are an acceptable alternative
to the 10 CFR Part 50, Appendix R fire protection features (69 FR
33536, June 16, 2004). The requirements in NFPA 805 address only
fire protection and the impacts of fire on the plant that have
already been evaluated. Based on this, the implementation of this
amendment does not create the possibility of a new or different kind
of accident from any kind of accident previously evaluated. The
proposed amendment does not introduce any new accident scenarios,
transient precursors, failure mechanisms, malfunctions, or limiting
single failures that could initiate a new accident. There will be no
adverse effect or challenges imposed on a safety related system as a
result of this proposed amendment.
Therefore, the possibility of a new or different kind of
accident from any kind of accident previously evaluated is not
created with the implementation of this amendment.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Operation of PINGP in accordance with the proposed amendment
does not involve a significant reduction in a margin of safety.
[[Page 19754]]
The proposed amendment 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 amendment does not
adversely affect existing plant safety margins or the reliability of
equipment assumed to mitigate accidents in the USAR. The proposed
amendment does not adversely affect the ability of SSCs to perform
their design function. SSCs required to safely shut down the reactor
and to maintain it in a safe shutdown condition remain capable of
performing their design function.
The purpose of this amendment is to permit PINGP to adopt a new
fire protection licensing basis which complies with the requirements
in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG
1.205. The NRC considers that NFPA 805 provides an acceptable
methodology and performance criteria for licensees to identify fire
protection systems and features that are an acceptable alternative
to the 10 CFR Part 50, Appendix R fire protection features (69 FR
33536; June 16, 2004). Engineering analyses, which may include
engineering evaluations, probabilistic safety assessments, and fire
modeling evaluations, have been performed to demonstrate that the
performance-based methods do not result in a significant reduction
in a margin of safety.
Based on this, the implementation of this amendment does not
significantly reduce a margin of safety. The proposed changes are
evaluated to ensure that the risk and safety margins are kept within
acceptable limits.
Therefore, the transition to NFPA 805 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 requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Robert D. Carlson.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, (SSES) Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment requests: December 19, 2012.
Description of amendment requests: The proposed amendments would
modify the SSES Unit 1 and SSES Unit 2 Technical Specifications (TS)
Section 2.1.1 to reflect a revised Low Pressure Safety Limit. The
change to TS Section 2.1.1 became necessary as a result of General
Electric (GE) PART 21 REPORT, SC05-03, ``Potential to Exceed Low
Pressure Technical Specification Safety Limit.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment changes the low pressure safety limit in
Technical Specification (TS) 2.1.1 from 785 psig [pounds per square
inch gauge] to 557 psig based on the capabilities of the current
critical power correlation used by Susquehanna (SPCB). The SPCB
correlation is approved for CPR [critical power ratio] calculations
by the NRC for reactor pressures > 571.4 psia [pounds per square
inch absolute] and is listed as an approved analytical method in TS
5.6.5.b.
The proposed changes will not alter existing Final Safety
Analysis Report (FSAR) design basis accident analysis assumptions,
add any accident initiators, or affect the function of the plant
safety-related structures, systems, or components (SSCs) as to how
they are operated, maintained, modified, tested, or inspected.
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 change to the Low Pressure Safety Limits does not result in
the need for any new or different FSAR design basis accident
analysis. The inclusion does not introduce new equipment that could
create a new or different kind of accident, and no new equipment
failure modes are created. In addition, the proposed change does not
affect the function of any safety-related SSC as to how they are
operated, maintained, modified, tested or inspected. As a result, no
new accident scenarios, failure mechanisms, or limiting single
failures are introduced as a result of this proposed amendment.
Therefore, the proposed amendment does not create a possibility
for an accident of a new or different type than those previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety is associated with the confidence in the
ability of the fission product barriers (i.e., fuel cladding,
reactor coolant pressure boundary, and containment structure) to
limit the level of radiation to the public. Evaluation of the 10 CFR
Part 21, ``Reporting of Defects and Noncompliance'' issue that
identified the need for the proposed change determined that there
was no decrease in the safety margin and therefore no threat to fuel
cladding integrity. The proposed changes to the Low Pressure Safety
Limits would not alter the way safety-related SSCs function and
would not alter the way PPL Susquehanna Units 1 and 2 are operated.
The proposed changes to the safety limit are within the capabilities
of the existing NRC approved CPR correlation and ensure valid CPR
calculations for the Anticipated Operational Occurrences (AOOs)
defined in the FSAR. The proposed amendment would have no impact on
the structural integrity of the fuel cladding, reactor coolant
pressure boundary, or containment structure. Based on the above
considerations, the proposed amendment would not degrade the
confidence in the ability of the fission product barriers to limit
the level of radiation to the public.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Meena K. Khanna.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit 1 and Unit 2 (Salem), Salem County, New Jersey
Date of amendment request: July 17, 2012, as supplemented on
January 28, 2013.
Description of amendment request: The proposed amendments would
revise Salem Technical Specifications (TS) 3.7.6.1 (Unit 1) and 3.7.6
(Unit 2), ``Control Room Emergency Air Conditioning System,'' to
eliminate the separate action statements for securing an inoperable
Control Area Air Conditioning System and Control Room Emergency Air
Conditioning System isolation damper in the closed position and
entering the actions for an inoperable control room envelope 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 Control Room Emergency Air Conditioning System (CREACS) is
not an initiator of or a precursor to any accident or transient. The
CREACS system is in standby during normal operation and initiates in
the
[[Page 19755]]
event of a safety injection signal or control room radiation
monitoring actuation in response to a design basis accident to
pressurize the Control Room Envelope (CRE) and provide filtration of
the CRE atmosphere to maintain the control room operator doses
within the limits of General Design Criteria (GDC) 19. The system
also operates in recirculation mode to mitigate the consequences of
a fire or toxic gas release that occurs outside of the CRE.
The design of plant equipment is not being modified by the
proposed amendment. The elimination of the action to secure the
isolation dampers between the normal Control Area Air Conditioning
System (CAACS) and the CREACS when these dampers are inoperable and
entering the actions for the inoperable control room boundary will
ensure operation of the plant within the limits of the radiological,
smoke and chemical hazard analyses. The intent of the original
action for securing the inoperable isolation damper in the closed
position was to maintain the boundary of the CRE. The actions for an
inoperable control room boundary ensure that mitigating actions are
implemented that maintain the CRE boundary within the limits of the
radiological, smoke and chemical hazard analyses.
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 to the TS to implement the actions for an
inoperable control room boundary when a normal CAACS and CREACS
isolation damper is inoperable do not introduce any new accident
precursors and do not involve any physical plant alterations or
changes in the methods governing normal plant operation that could
initiate a new or different kind of accident. The proposed amendment
does not alter the function of the system to initiate and pressurize
the control room envelope in the event of a DBA nor alter the
ability to initiate CREACS in the recirculation mode in response to
a fire or chemical release that occurs outside of the CRE.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is related to the ability of the fission
product barriers (fuel cladding, reactor coolant system, and primary
containment) to perform their design functions during and following
postulated accidents. The proposed amendment does not alter
setpoints or limits established or assumed by the accident analyses.
The control room envelope is considered a barrier for the control
room operators during a design basis accident radiological release
and a barrier in the event of a fire or chemical hazard that occurs
outside of the CRE. Implementing the actions for an inoperable
control room boundary in the event of an inoperable isolation damper
between the normal CAACS and CREACS ensure operation of the plant
within the limits of the radiological, smoke and chemical hazard
analysis. The actions for an inoperable control room boundary ensure
that mitigating actions are implemented that maintain the CRE
boundary within the limits of the radiological, smoke and chemical
hazard analyses. Therefore the plant will continue to be operated
consistent with the plant safety analyses.
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 requests involve no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Meena K. Khanna.
South Carolina Electric and Gas 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: March 13, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3 in regard to the Chemical and Volume
Control System (CVS) by: (1) Providing a spring-assisted check valve
around the air-operated Reactor coolant System (RCS) Purification
Return Line Stop Check Valve, (2) replacing the CVS zinc addition
inboard containment isolation lift check valve with an air-operated
globe valve and a thermal relief valve and (3) separating the zinc and
hydrogen injection paths and relocate the zinc injection path.
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 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 changes to provide a spring-assisted check valve located in
the bypass line around the makeup stop check valve would continue to
meet the existing design functions because the ASME Boiler and
Pressure Vessel Code (ASME Code) Section III valves will maintain
the flow isolation design function and preserve the Reactor Coolant
System (RCS) pressure boundary safety function. The replacement of
the Chemical and Volume Control System (CVS) zinc addition inboard
containment isolation lift check valve with an air operated globe
valve and addition of a pressure relief valve would continue to meet
the containment isolation and RCS pressure boundary design functions
because the replacement valves will be designed, analyzed, tested
and qualified, including seismic qualification, to ASME Code Section
III requirements. Separating the zinc and hydrogen injection paths
and relocating the zinc injection point would continue to meet
containment boundary requirements, including containment isolation
and in-service testing, and preserve the RCS pressure boundary
safety functions because the revised containment isolation
configuration is consistent with those described in 10 CFR Part 50,
Appendix A, General Design Criterion (GDC) 55, and the additional
valves and piping will be qualified to ASME Code Section III.
Because the proposed CVS changes would preserve the CVS safety-
related design functions, the probability of an accident previously
evaluated is not affected.
The CVS safety functions have been preserved, because the
proposed CVS configuration changes, including revised valve types,
will perform the same safety functions as the current design. The
proposed CVS configuration changes would neither impact any accident
source term parameter or fission product barrier nor affect
radiological dose consequence analysis.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The additional containment penetration is similar in form, fit,
and function to the CVS combined zinc/hydrogen containment
penetration that is currently described in the Updated Final Safety
Analysis Report. Because the CVS changes use valve types, piping,
and a containment penetration consistent with those already
described in the Updated Final Safety Analysis Report, no new
failure modes or equipment failure initiators are introduced by
these changes. Accordingly, the proposed changes do not create any
new malfunctions, failure mechanisms, or accident initiators.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
[[Page 19756]]
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The containment isolation and pressure relief functions would
not be changed by this activity and are consistent with the existing
design. The proposed CVS containment penetration is similar in form,
fit, and function to existing CVS combined zinc/hydrogen containment
penetration and, therefore, does not affect containment or its
ability to perform its design function. The addition of these CVS
components, including piping, a spring-assisted check valve, an air-
operated containment isolation valve, a thermal relief valve and the
additional CVS containment penetration do not impact a design basis
or safety limit. Because the CVS design functions of controlling the
RCS oxygen concentration, reducing radiation fields, containment
isolation and overpressure protection within existing limits are not
changed by this activity and are bounded by the existing design,
there is no change to any current margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.
NRC Acting Branch Chief: Lawrence Burkhart.
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: February 15, 2013.
Description of amendment request: The proposed change would amend
Combined Licenses Nos. NPF-91 and NPF-92 for Vogtle Electric Generating
Plant (VEGP) Units 3 and 4 by departing from the plant-specific design
control document Tier 2* material by revising reference document APP-
OCS-GEH-320, ``AP1000 Human Factors Engineering Integrated System
Validation Plan'' from Revision D to Revision 2. APP-OCS-GEH-320 is
incorporated by reference in the updated final safety analysis report
(UFSAR) as a means to implement the activities associated with the
human factors engineering verification and validation.
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 Integrated System Validation (ISV) provides a comprehensive
human performance-based assessment of the design of the AP1000
Human-System Interface (HSI) resources, based on their realistic
operation within a simulator-driven Main Control Room (MCR). The ISV
is part of the overall AP1000 Human Factors Engineering (HFE)
program. The changes are to the ISV Plan to clarify the scope and
amend the details of the methodology. The ISV Plan is needed to
perform, in the simulator, the scenarios described in the document.
The functions and tasks allocated to plant personnel can still be
accomplished after the proposed changes. The performance of the
tests governed by the ISV Plan provides additional assurances that
the operators can appropriately respond to plant transients. The ISV
Plan does not affect the plant itself. Changing the ISV Plan does
not affect prevention and mitigation of abnormal events, e.g.,
accidents, anticipated operational occurrences, earthquakes, floods
and turbine missiles, or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely
affected. The changes do 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 changes do not involve any safety-related SSC
or 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.
The changes to the ISV Plan affect the testing and validation of
the Main Control Room and Human System Interface using a plant
simulator. Therefore, the changes do not affect the safety-related
equipment itself, nor do they affect equipment which, if it failed,
could initiate an accident or a failure of a fission product
barrier. No analysis is adversely affected. No system or design
function or equipment qualification will be adversely affected by
the changes. This activity will not allow for a new fission product
release path, nor will it result in a new fission product barrier
failure mode, nor create a new sequence of events that would result
in significant fuel cladding failures. In addition, the changes do
not result in a new failure mode, malfunction or sequence of events
that could affect safety or safety-related equipment.
Therefore, this activity does not create the possibility of a
new or different kind of accident than any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to the ISV Plan affect the testing and validation of
the Main Control Room and Human System Interface using a plant
simulator. Therefore, the changes do not affect the assessments or
the plant itself. These changes do not affect safety-related
equipment or equipment whose failure could initiate an accident, nor
does it adversely interface with safety-related equipment or fission
product barriers. No safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by the requested change.
Therefore, there is no 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 Acting Branch Chief: Lawrence Burkhart.
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
[[Page 19757]]
under the special circumstances provision in 10 CFR 51.22(b) and has
made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: November 14, 2012.
Brief description of amendments: The amendments relocate the
Technical Specification (TS) requirements for motor-operated valve
thermal overload protection from the TSs to the Technical Requirements
Manual.
Date of issuance: March 19, 2013.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 209 and 170.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the License and TSs.
Date of initial notice in Federal Register: January 8, 2013 (78 FR
1270).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 2013.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant (PNPP), Unit 1, Lake County, Ohio
Date of application for amendment: July 3, 2012, supplemented by
letter dated January 7, 2013.
Brief description of amendment: The proposed amendment would modify
PNPP's Technical Specifications (TS) 3.8.1, ``AC [alternating current]
Sources--Operating.'' Specifically, the proposed amendment will modify
nine surveillance requirements (SRs) by excluding Division 3 from the
current mode restrictions, thus allowing performance of the subject SRs
in any mode of plant operation. The proposed amendment also deletes
expired TS 3.8.1 provisions regarding use of a delayed access circuit.
Date of issuance: March 5, 2013.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 162.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: November 13, 2012 (77
FR 67682). The January 7, 2013 supplement contained clarifying
information and did not change the NRC staff's initial proposed finding
of no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2013.
No significant hazards consideration comments received: No.
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 18, 2013.
Brief description of amendment: The proposed amendment would depart
from VEGP Units 3 and 4 plant-specific Design Control Document (DCD)
Tier 2 material incorporated into the Updated Final Safety Analysis
Report (UFSAR) by revising the structural criteria code for anchoring
of headed shear reinforcement bar within the nuclear island basemat.
Date of issuance: March 1, 2013.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: Unit 3-5, and Unit 4-5.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: January 29, 2013 (78 FR
6142).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 25th day of March 2013.
For The Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2013-07467 Filed 4-1-13; 8:45 am]
BILLING CODE 7590-01-P