[Federal Register Volume 75, Number 45 (Tuesday, March 9, 2010)]
[Notices]
[Pages 10823-10833]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-4523]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0081]
Biweekly Notice; Applications and Amendments to Facility
Operating 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 (the
Commission or 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
upon a determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 11, 2010, to February 24, 2010. The
last biweekly notice was published on February 23, 2010 (75 FR 8139).
Notice of Consideration of Issuance of Amendments to Facility Operating
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 Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
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.
Written comments may be submitted by mail to the Chief, Rulemaking
and Directives Branch (RDB), TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be faxed to the RDB at 301-492-3446. Documents may be examined, and/or
copied for a fee, at the NRC's Public Document Room (PDR), located at
One White Flint North, Public File Area O1F21, 11555 Rockville Pike
(first floor), Rockville, Maryland.
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
[[Page 10824]]
subject facility operating license. Requests for a hearing and a
petition for leave to intervene shall be filed in accordance with the
Commission's ``Rules of Practice for Domestic Licensing Proceedings''
in 10 CFR part 2. Interested person(s) should consult a current copy of
10 CFR 2.309, which is available at the Commission's PDR, located at
One White Flint North, Public File Area O1F21, 11555 Rockville Pike
(first floor), Rockville, Maryland. Publicly available records will be
accessible from the Agencywide Documents Access and Management System's
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web
site, 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, 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 E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the participant should
contact the Office of the Secretary by e-mail at
[email protected], or by telephone at (301) 415-1677, to request
(1) a digital 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 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 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 EIE, users will be required to install a Web
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC 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
[[Page 10825]]
the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC 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 Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (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
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, 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. 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. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, 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 PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 28, 2009.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.8.1, ``AC Sources-Operating,'' to
restrict voltage limits for the applicable TS 3.8.1 surveillances
governing the Emergency Diesel Generators (EDGs).
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?
No. The increase in the minimum EDG output voltage acceptance
value in TS 3.8.1 Surveillance Requirements does not adversely
affect any of the parameters in the accident analyses. The proposed
change increases the minimum allowed EDG output voltage to ensure
that sufficient voltage is available to operate the required
Emergency Safety Feature (ESF) equipment under accident conditions.
Additionally the increase in minimum voltage output voltage allowed
ensures that adequate voltage is available to support the
assumptions made in the Design Bases Accident (DBA) analyses. This
conservative change of the EDG voltage output acceptance criteria
does not affect the probability of evaluated accidents, but rather
provides increased assurance that the EDGs will provide a sufficient
voltage. 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?
No. The increase in the minimum EDG output voltage acceptance
criterion supports the assumptions in the accident analyses that
sufficient voltage will be available to operate ESF equipment on the
Class 1E buses when these buses are powered from the Emergency
Diesel Generators. The maximum EDG output voltage of 4580 volts is
not affected by this change. The change in minimum output voltage
from 3740 to 3950 volts ensures the reliability of the onsite
emergency power source. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed change involve a significant reduction in
margin of safety?
This proposed license amendment is limited to increasing the
minimum EDG output voltage acceptance criterion in TS 3.8.1
Surveillance Requirements. No other surveillance criterion is
affected. The surveillance frequencies and test requirement are
unchanged. This amendment provides increased assurance that the EDG
will provide sufficient voltage to its respective components to
ensure design requirements are satisfied. Therefore, the proposed
change will 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. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
[[Page 10826]]
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: July 1, 2009.
Description of amendment request: The proposed amendments would
revise TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation'' and TS
1.1, ``Definitions.'' The proposed amendments support plant
modifications which would replace the existing Source Range (SR) and
Intermediate Range (IR) excore detector systems with equivalent neutron
monitoring systems. The new instrumentation will perform both the SR
and the IR monitoring functions.
Implementation of the above changes will entail plant modifications
and will impact the Updated Final Safety Analysis Reports (UFSAR). The
necessary UFSAR revisions will be submitted in accordance with 10 CFR
50.71(e).
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 Technical Specification changes are in support of a
plant modification involving the replacement and upgrade of the
Nuclear Instrumentation System (NIS) Source Range and Intermediate
Range instrumentation. The specific Technical Specification changes
are associated with 1) the methods of calibrating NIS channels; 2)
the definition of Nominal Trip Setpoint; 3) the specific Nominal
Trip Setpoint and Allowable Values for various NIS channels,
including the Intermediate Range, Source Range and Intermediate
Range Permissive ``P-6'' instrumentation; 4) the addition of
specific requirements to be taken if an as-found Intermediate Range
or Source Range channel setpoint is outside its predefined as-found
tolerance; and 5) the addition of specific requirements regarding
resetting of an Intermediate Range or Source Range channel setpoint
within an as-left tolerance.
The NIS is accident mitigation equipment and does not affect the
probability of any accident being initiated. In addition, none of
the above-mentioned proposed Technical Specification changes affect
the probability of any accident being initiated.
The performance of the replacement SR and IR detectors and
associated equipment will equal or exceed that of the existing
instrumentation. The proposed changes to Nominal Trip Setpoints and
Allowable Values are based on accepted industry standards and will
preserve assumptions in the applicable accident analyses. None of
the proposed changes alter any assumption previously made in the
radiological consequences evaluations, nor do they affect mitigation
of the radiological consequences of an accident previously
evaluated.
In summary, the proposed changes 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
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of any of the proposed changes.
The NIS is not capable by itself of initiating any accident. Other
than the replacement of the detectors themselves and the associated
hardware, no physical changes to the overall plant are being
proposed. No changes to the overall manner in which the plant is
operated are being proposed. Therefore, none of the proposed changes
will 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 related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant system
pressure boundary, and the containment barriers. The modification to
replace the SR and IR detectors and associated equipment will not
have any impact on these barriers. In addition, the proposed
Technical Specification changes will not have any impact on these
barriers. No accident mitigating equipment will be adversely
impacted as a result of the modification. The proposed changes do
not affect any safety analysis conclusions because the SR and IR
neutron flux trips are not explicitly credited in any accident
analysis. The replacement instrumentation will have overall
performance capabilities equal to or greater than those for the
existing instrumentation. Therefore, existing safety margins will be
preserved. None of the proposed changes will 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. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: July 1, 2009.
Description of amendment request: The proposed amendments would
revise TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation.'' The
proposed amendments support plant modifications which would replace the
existing Source Range (SR) and Intermediate Range (IR) excore detector
systems with equivalent neutron monitoring systems. The new
instrumentation will perform both the SR and the IR monitoring
functions.
Implementation of the above changes will entail plant modifications
and will impact the Updated Final Safety Analysis Reports (UFSAR). The
necessary UFSAR revisions will be submitted in accordance with 10 CFR
50.71(e).
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 Technical Specification changes are in support of a
plant modification involving the replacement and upgrade of the
Nuclear Instrumentation System (NIS) Source Range and Intermediate
Range instrumentation. The specific Technical Specification changes
are associated with (1) the methods of calibrating NIS channels; (2)
the definition of Nominal Trip Setpoint; (3) the specific Nominal
Trip Setpoint and Allowable Values for various NIS channels,
including the Intermediate Range, Source Range and Intermediate
Range Permissive ``P-6'' instrumentation; (4) the addition of
specific requirements to be taken if an as-found Intermediate Range
or Source Range channel setpoint is outside its predefined as-found
tolerance; and (5) the addition of specific requirements regarding
resetting of an Intermediate Range or Source Range channel setpoint
within an as-left tolerance.
The NIS is accident mitigation equipment and does not affect the
probability of any accident being initiated. In addition, none of
the above-mentioned proposed Technical Specification changes affect
the probability of any accident being initiated.
The performance of the replacement SR and IR detectors and
associated equipment will equal or exceed that of the existing
instrumentation. The proposed changes to Nominal Trip Setpoints and
Allowable Values are based on accepted industry standards and will
preserve assumptions in
[[Page 10827]]
the applicable accident analyses. None of the proposed changes alter
any assumption previously made in the radiological consequences
evaluations, nor do they affect mitigation of the radiological
consequences of an accident previously evaluated.
In summary, the proposed changes 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
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of any of the proposed changes.
The NIS is not capable by itself of initiating any accident. Other
than the replacement of the detectors themselves and the associated
hardware, no physical changes to the overall plant are being
proposed. No changes to the overall manner in which the plant is
operated are being proposed. Therefore, none of the proposed changes
will 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 related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant system
pressure boundary, and the containment barriers. The modification to
replace the SR and IR detectors and associated equipment will not
have any impact on these barriers. In addition, the proposed
Technical Specification changes will not have any impact on these
barriers. No accident mitigating equipment will be adversely
impacted as a result of the modification. The proposed changes do
not affect any safety analysis conclusions because the SR and IR
neutron flux trips are not explicitly credited in any accident
analysis. The replacement instrumentation will have overall
performance capabilities equal to or greater than those for the
existing instrumentation. Therefore, existing safety margins will be
preserved. None of the proposed changes will 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. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina; Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units
1 and 2, Mecklenburg County, North Carolina; Docket Nos. 50-413 and 50-
414, Catawba Nuclear Station, Units 1 and 2, York County, South
Carolina
Date of amendment request: May 18, 2009.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to adopt Technical Specification
Task Force (TSTF) Standard Technical Specification Change Traveler
TSTF-248. TSTF 248 modifies the definition of shutdown margin.
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:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The revision to SDM [shutdown margin] definition will result in
analytical flexibility for determining SDM. Changes in the
definition will not have an impact on the probability of an accident
previously evaluated.
The introduction of this definition change does not change
continued compliance with all applicable regulatory requirements and
design criteria (e.g., train separation, redundancy, and single
failure). Therefore, since all plant systems will continue to
function as designed, all plant parameters will remain within their
design limits. As a result, the proposed changes will not increase
the consequences of an accident.
Based on this discussion, the proposed amendments do not
significantly increase the probability or consequences of an
accident previously evaluated.
Criterion 2:
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Revising the TS [Technical Specifications] definition of SDM
would not require core designers to revise any SDM boron
calculations. Rather, it would afford the analytical flexibility for
determining SDM for a particular circumstance.
The proposed changes do not involve any change in the design,
configuration, or operation of the nuclear plant. The current plant
safety analyses, therefore, remain complete and accurate in
addressing the design basis events and in analyzing plant response
and consequences.
The Limiting Conditions for Operations, Limiting Safety System
Settings and Safety Limits specified in the Technical Specifications
are not affected by the proposed changes. As such, the plant
conditions for which the design basis accident analyses were
performed remain valid.
The amendment does not introduce a new mode of plant operation
or new accident precursors, does not involve any physical
alterations to plant configurations or make changes to system set
points that could initiate a new or different kind of accident.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3:
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their accident mitigation
functions. These barriers include the fuel and fuel cladding, the
reactor coolant system, and the containment and containment related
systems. The proposed changes will not impact the reliability of
these barriers to function. Radiological doses to plant operators or
to the public will not be impacted as a result of the proposed
change. The change in the TS definition will have no impact to these
barriers. Adequate SDM will continue to be ensured for all
operational conditions.
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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: August 6, 2009.
Description of amendment request: The proposed amendments would
revise the Technical Specifications by changing the surveillance
requirement for the low temperature overpressure protection system
(LTOP) from 6 months to 18 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 10828]]
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?
No. This is a revision to the Technical Specification (TS)
Surveillance Requirement (SR) for performing the channel calibration
for the power operated relief valve (PORV). As such, the TS SR
interval extension continues to ensure the calibration is performed
in a time frame supported by current analysis. The instrumentation
loop has been upgraded to an environmentally qualified
instrumentation loop with improved instrument uncertainty and
reliability. The accidents previously evaluated have not changed.
Therefore, extending the TS SR frequency from 6 months to 18
months does not significantly increase 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?
No. This revision does not impact the LTOP evaluation analysis.
The method for testing remains the same. The proposed SR frequency
is supported by an environmentally qualified instrumentation loop
with improved instrument uncertainty and reliability.
Therefore, extending the TS SR frequency from 6 months to 18
months will not create the possibility of a new or different kind of
accident from any kind of accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The proposed change does not adversely affect any plant
safety limits, setpoints, or design parameters. The change also does
not adversely affect the fuel, fuel cladding, Reactor Coolant
System, or Containment Operability.
Therefore, extending the TS SR frequency from 6 months to 18
months 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. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: August 31, 2009.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to allow one of the two required
230kV switchyard 125 VDC power source batteries to be inoperable for up
to 10 hours for the purpose of replacing an entire battery bank and
performing the required testing.
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?
No. This License Amendment Request (LAR) proposes to permit one
of the two 230 kV switchyard 125 VDC batteries to be out of service
for up to ten days when it is necessary to replace and test a
complete battery (all cells of one battery bank). The capacity of
each battery, needing only 58 of 60 cells to be available (i.e., two
cells can be jumpered out), is sufficient to carry the loads of both
distribution centers during replacement.
The 230kV switchyard 125 VDC power system is credited to provide
uninterruptible power to specified loads during certain design basis
events. The probability of any of these events occurring is not
impacted by removing one of the batteries for replacement. The
consequences associated with permitting a 230 kV switchyard 125 VDC
battery to be out of service for up to ten days for battery
replacement have been evaluated. The likelihood of an event
occurring during the additional time a battery bank will be out of
service is essentially the same as that of an event occurring during
the 24 hour period permitted by the existing completion time.
Operation in accordance with the amendment authorizing this change
would not involve any accident initiation sequences or radiological
release pathways that could affect the consequences of any accident
analyzed. Use of this additional time for battery replacement will
be infrequent since battery replacement normally is performed at or
near the end of the twenty year qualified life.
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?
No. This License Amendment Request (LAR) proposes to permit one
of the two 230 kV switchyard 125 VDC batteries to be out of service
for up to ten days when it is necessary to replace and test a
complete battery (all cells of one battery). Operation in accordance
with this proposed amendment will not result in any new plant
equipment, alter the present plant configuration, nor adversely
affect how the plant is currently operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
No. This License Amendment Request (LAR) proposes to permit one
of the two 230 kV switchyard 125 VDC batteries to be out of service
for up to ten days when it is necessary to replace and test a
complete battery (all cells of one battery).
Since the proposed change will not physically alter the present
plant configuration nor adversely affect how the plant is currently
operated, the proposed change does not adversely affect any plant
safety limits, setpoints, or design parameters. The change also does
not adversely affect the fuel, fuel cladding, Reactor Coolant System
or containment integrity.
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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina; Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units
1 and 2, Mecklenburg County, North Carolina; Docket Nos. 50-413 and 50-
414, Catawba Nuclear Station, Units 1 and 2, York County, South
Carolina
Date of amendment request: September 30, 2009.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to allow performance of testing
containment spray nozzles for nozzle blockage following activities
which could result in nozzle blockage, rather than a fixed periodic
basis. Currently the testing for nozzle blockage is performed every 10
years.
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:]
Does the proposed change involve a significant increase in the
probability or
[[Page 10829]]
consequences of an accident previously evaluated?
No. The proposed amendment will modify CNS [Catawba Nuclear
Station] SR [surveillance requirement] 3.6.6.7, MNS [McGuire Nuclear
Station] SR 3.6.6.7, and ONS [Oconee Nuclear Station] SR 3.6.5.8 to
change the frequency for verifying spray nozzles are unobstructed.
The proposed change modifies the frequency for performance of a
surveillance test which does not impact any failure modes that could
lead to an accident. The proposed frequency change does not affect
the ability of the spray nozzles or spray system to perform its
accident mitigation function as assumed and therefore there is no
effect on the consequence of any accident. Verification of no
blockage continues to be required, but now verification will be
performed following activities that could result in nozzle blockage.
Based on this discussion, the proposed amendment does not increase
the probability or consequence of an accident previously evaluated.
[Criterion 2:]
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed amendment will modify CNS SR 3.6.6.7, MNS SR
3.6.6.7, and ONS SR 3.6.5.8 to change the frequency for verifying
spray nozzles are unobstructed. The spray systems are not being
physically modified and there is no impact on the capability of the
system to perform accident mitigation functions. No system setpoints
are being modified and no changes are being made to the method in
which borated water is delivered to the spray nozzles. The testing
requirements imposed by this proposed change to check for nozzle
blockage following activities that could cause nozzle blockage do
not introduce new failure modes for the system. The proposed
amendment does not introduce accident initiators or malfunctions
that would cause a new or different kind of accident. Therefore, the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
[Criterion 3:]
Does the proposed amendment involve a significant reduction in a
margin of safety?
No. The proposed amendment will modify CNS SR 3.6.6.7, MNS SR
3.6.6.7, and ONS SR 3.6.5.8 to change the frequency for verifying
spray nozzles are unobstructed. The proposed change does not change
or introduce any new setpoints at which mitigating functions are
initiated. No changes to the design parameters of the spray systems
are being proposed. There are no changes in system operation being
proposed by this change that would impact an established safety
margin. The proposed change modifies the frequency for verification
of nozzle operability in such a way that continued high confidence
exists that the spray systems will continue to function as designed.
Therefore, based on the above, the proposed amendment does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: November 19, 2009, as supplemented by
letter dated January 28, 2010.
Description of amendment request: The proposed change will modify
the test acceptance criteria in Surveillance Requirement (SR) 3.8.1.10
for the Diesel Generator endurance surveillance test. The proposed
change will also incorporate changes to the Standard Technical
Specifications made by Technical Specification Task Force (TSTF) 238-A,
Revision 3 and TSTF-276-A, Revision 2. Specifically, the proposed
change will modify SR notes in TS 3.8.1 and TS 3.8.4
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?
No. The proposed changes revise the acceptance criteria to be
applied to an existing surveillance test of the facility emergency
diesel generators (EDGs), allows deviation from that acceptance
criteria for certain grid conditions, and allows testing in modes
that is normally not done. Performing a surveillance test is done
under conditions where it is not an accident initiator and does not
increase the probability of an accident occurring. The proposed new
acceptance criteria will assure that the EDGs are capable of
carrying the peak electrical loading assumed in the various existing
safety analyses which take credit for the operation of the EDGs.
Establishing acceptance criteria that bound existing analyses
validates the related assumption used in those analyses regarding
the capability of equipment to mitigate accident conditions. The
deviation allowed for grid conditions does not affect the capability
of the testing to achieve these purposes. The proposed change to
allow testing in modes normally restricted requires an evaluation to
ensure, prior to performing the test, that the potential
consequences are capable of being addressed by existing procedures
and does not create transients or conditions that could
significantly affect the possibility of an accident. 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 or different
kind of accident from any accident previously evaluated?
No. The proposed changes revise the test acceptance criteria for
a specific performance test conducted on the existing EDG, allows
deviation from that acceptance criteria for certain grid conditions,
and allows testing in modes that is normally not done. The proposed
changes do not involve installation of new equipment or modification
of existing equipment, so no new equipment failure modes are
introduced. The proposed revision to the EDG surveillance test
acceptance criteria also is not a change to the way that the
equipment or facility is operated and no new accident initiators are
created. The proposed testing on line must be evaluated to assure
plant safety is maintained or enhanced, inherent in such an
evaluation would be that the testing does not create the possibility
of a new or different kind of accident. 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?
No. The conduct of performance tests on safety-related plant
equipment is a means of assuring that the equipment is capable of
maintaining the margin of safety established in the safety analyses
for the facility. The proposed change in the EDG technical
specification surveillance test acceptance criteria is consistent
with values assumed in existing safety analyses and is consistent
with the design rating of the EDGs. The allowance for certain grid
conditions does not alter this conclusion since the power factors
are conservatively determined. Testing allowed in modes when it is
not normally performed is limited to conditions where an evaluation
is performed to assure plant safety is maintained or enhanced.
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 Branch Chief: Nancy L. Salgado.
[[Page 10830]]
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: December 18, 2009.
Description of amendment request: The proposed amendment would
incorporate the use of alternate methodologies for the calculation of
reactor pressure vessel beltline weld initial reference temperatures,
the calculation of the adjusted reference temperatures (ARTs), the
development of the reactor pressure vessel pressure-temperature (P-T)
limit curves, and the low temperature reactor coolant system (RCS)
overpressure analysis into Technical Specification (TS) 5.6.4. The
amendment would also revise the analysis requirement for the low
temperature RCS overpressure events from 21 to 32 Effective Full Power
Years (EFPY) contained in Operating License (OL) Condition 2.C(3)(d).
An application that addressed similar issues was previously submitted
on April 15, 2009, and the notice of that application was provided in
the Federal Register on June 16, 2009 (72 FR 28577). Since the licensee
eliminated one of the alternate methodologies for the calculation of
the adjusted reference temperature (as described in the April 15, 2009,
application) and replacing it with the existing Nuclear Regulatory
Commission (NRC)-approved methodology, which is described in Regulatory
Guide 1.99, Revision 2, ``Radiation Embrittlement of Reactor Vessel
Materials'', in December 19, 2009, the application is being renoticed
in its entirety. The notice supersedes the notice published in the
Federal Register on June 16, 2009.
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 amendment request proposes two changes to the TS/OL. The
first change incorporates the use of alternative methodologies to
develop the [Davis-Besse Nuclear Power Station, Unit No. 1] DBNPS P-
T limit curves and [low temperature over pressure] LTOP limits into
TS 5.6.4 to augment the existing listed methodology of BAW-10046A,
Revision 2. The second change revises OL Condition 2.C(3)(d) to
reflect the revised LTOP analysis is valid to 32 [Effective Full
Power Years] EFPY.
The first change incorporates the use of Topical Report BAW-
2308, Revisions 1-A and 2-A and [American Society of Mechanical
Engineers] ASME Code Cases N-588 and N-640. The topical report and
ASME code cases have been approved or accepted for use by the NRC
(provided that any conditions/limitations are satisfied). The
proposed additions to the methodologies for the reactor vessel P-T
curve and LTOP limit development provide an acceptable means of
satisfying the requirements of 10 CFR 50, Appendix G. The proposed
additions do not alter the design, function, or any operation of any
plant equipment. Therefore, the proposed additions do not affect the
probability or consequences of any previously evaluated accidents,
including reactor coolant pressure boundary failures.
The second change is considered administrative in nature and
reflects the revised methodologies. It will not alter the design,
function, or operation of any plant equipment. Therefore, the
proposed change does not affect the probability or consequences of
any previously evaluated accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The amendment request proposes two changes to the TS/OL. The
first change incorporates the use of alternative methodologies to
develop the DBNPS P-T limit curves and LTOP limits into TS 5.6.4 to
augment the existing listed methodology of BAW-10046A, Revision 2.
The second change revises OL Condition 2.C(3)(d) to reflect that the
revised analysis is valid to 32 EFPY.
The first change incorporates methodologies that either have
been approved or accepted for use by the NRC (provided that any
conditions/limitations are satisfied). The changes do not alter the
design, function, or operation of any plant equipment. The P-T limit
curves and LTOP limits will provide the same level of protection to
the reactor coolant boundary as was previously evaluated. Therefore,
the proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The second change is considered administrative in nature and
reflects the revised methodologies. It will not alter the design or
operation of any plant equipment. 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 amendment request proposes two changes to the TS/OL. The
first change incorporates the use of alternative methodologies to
develop the DBNPS P-T limit curves and LTOP limits into TS 5.6.4 to
augment the existing listed methodology of BAW-10046A, Revision 2.
The second change revises OL Condition 2.C(3)(d) to reflect that the
revised analysis is valid to 32 EFPY. The first change incorporates
methodologies that either have been approved or accepted for use by
the NRC (provided that any conditions/limitations are satisfied).
The second change is considered administrative in nature and
reflects the revised methodologies. The changes do not alter the
design, function, or operation of any plant equipment. The P-T limit
curves and LTOP limits will provide the same level of protection to
the reactor coolant boundary as was previously evaluated. 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Stephen Campbell.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: November 30, 2009.
Description of amendment request: The proposed amendment would
modify conditions and associated actions to Technical Specification
3.8.1, ``AC [Alternating Current] Sources Operating.'' The proposed
amendment would revise the Completion Time for restoring one or more
inoperable diesel generators (DGs) in one train to an operable status
and increase the Completion Time for confirming that the other DGs are
not impacted by a common cause failure. 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 diesel generators (DGs) are designed as backup alternating
current (ac) power sources in the event of loss of offsite power.
The proposed changes to Completion Times associated with determining
inoperable DGs are not subject to common cause failure and
restoration of inoperable DGs and the deletion of the note
referencing the C-S DG do not change the conditions, operating
configurations, or minimum amount of operating equipment assumed in
the safety analysis accident mitigation. No changes are proposed in
the manner in which the DGs provide plant protection.
[[Page 10831]]
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 associated with determining inoperable DGs
are not subject to common cause failure and restoration of
inoperable DGs and the deletion of the note referencing the C-S DG
do not involve a change in design, configuration, or method of
operation of the plant. The proposed changes will not alter the
manner in which equipment operation is initiated, nor will the
functional demands on credited equipment be changed. The capability
of the DGs to perform their required safety function will not be
affected. The proposed changes do not affect the interaction of the
DGs with any system whose failure or malfunction can initiate an
accident. As such, no new failure modes are being introduced.
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 DGs are designed as backup AC power sources in the event of
loss of offsite power. The proposed changes associated with
determining inoperable DGs are not subject to common cause failure
and restoration of inoperable DGs and the deletion of the note
referencing the C-S DG do not change conditions, operating
configurations, or minimum amount of operating equipment assumed in
the safety analysis accident mitigation. The proposed changes do not
alter the plant design, including instrument setpoints, nor do they
alter the assumptions contained in the safety analyses. No changes
are proposed in the manner in which the DGs provide plant protection
or which create new modes of plant operation.
Therefore, the change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 16, 2009.
Description of amendment request: The proposed change would revise
the approved fire protection program as described in the Wolf Creek
Generating Station (WCGS) Updated Safety Analysis Report (USAR) to
allow use of the fire-resistive cable for certain power and control
cables associated with two motor-operated valves on Train B Component
Cooling Water System. This will be a deviation from certain technical
commitments to Title 10 of the Code of Federal Regulations (10 CFR)
Part 50, Appendix R, Section III.G.2, as described in Appendix 9.5E of
the WCGS USAR.
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 structures, systems and components are
not impacted by the proposed change. The proposed change involves
the use of fire-resistive cable at WCGS for certain power and
control cables associated with two motor-operated valves (EGHV0016
and EGHV0054) on Train B Component Cooling Water System and will not
initiate an event. The proposed change does not alter or prevent the
ability of structures, systems, and components (SSCs) from
performing their intended function to mitigate the consequences of
an initiating event within the assumed acceptance limits. The
Meggitt Si 2400 fire-resistive cable has been independently tested
to applicable requirements and the implementation design reflects
the test results. Therefore, the probability of any accident
previously evaluated is not increased. Equipment required to
mitigate an accident remains capable of performing the assumed
function.
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 will not alter the requirements or function
for systems required during accident conditions. The design function
of structures, systems and components are not impacted by the
proposed change. No new or different accidents result from
implementing Meggitt Si 2400 fire-resistive cable in Fire Areas A-16
and A-21. The Meggitt Si 2400 fire-resistive cable has been
independently tested to applicable requirements and the
implementation design reflects the test results. The use of Meggitt
Si 2400 fire-resistive cable is not a significant change in the
methods governing normal plant operation.
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 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 for
an unacceptable period of time without mitigating actions. The
proposed change does not affect systems that respond to safely shut
down the plant and to maintain the plant in a safe shutdown
condition.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 16, 2009.
Description of amendment request: The proposed change would revise
the approved fire protection program as described in the Wolf Creek
Generating Station (WCGS) Updated Safety Analysis Report (USAR) to
allow use of the fire-resistive cable for certain power and control
cables associated with two motor-operated valves on Train B Component
Cooling Water System. This will be a deviation from certain technical
commitments to Title 10 of the Code of Federal Regulations (10 CFR)
Part 50, Appendix R, Section III.G.2, as described in Appendix 9.5E of
the WCGS USAR.
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?
[[Page 10832]]
Response: No.
The design function of structures, systems and components are
not impacted by the proposed change. The proposed change involves
the use of fire-resistive cable at WCGS for certain power and
control cables associated with two motor-operated valves (EGHV0016
and EGHV0054) on Train B Component Cooling Water System and will not
initiate an event. The proposed change does not alter or prevent the
ability of structures, systems, and components (SSCs) from
performing their intended function to mitigate the consequences of
an initiating event within the assumed acceptance limits. The
Meggitt Si 2400 fire-resistive cable has been independently tested
to applicable requirements and the implementation design reflects
the test results. Therefore, the probability of any accident
previously evaluated is not increased. Equipment required to
mitigate an accident remains capable of performing the assumed
function.
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 will not alter the requirements or function
for systems required during accident conditions. The design function
of structures, systems and components are not impacted by the
proposed change. No new or different accidents result from
implementing Meggitt Si 2400 fire-resistive cable in Fire Areas A-16
and A-21. The Meggitt Si 2400 fire-resistive cable has been
independently tested to applicable requirements and the
implementation design reflects the test results. The use of Meggitt
Si 2400 fire-resistive cable is not a significant change in the
methods governing normal plant operation.
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 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 for
an unacceptable period of time without mitigating actions. The
proposed change does not affect systems that respond to safely
shutdown the plant and to maintain the plant in a safe shutdown
condition.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Notice of Issuance of Amendments to Facility Operating 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.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, 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 are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management System (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, 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 Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: March 29, 2009, as supplemented
by letters dated September 21 and December 22, 2009.
Brief description of amendment: The amendment established a more
restrictive acceptance criterion for surveillance requirement (SR)
3.8.6.6 regarding periodic verification of capacity for the affected
station batteries.
Date of issuance: February 24, 2010.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 264.
Facility Operating License No. DPR-26: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: May 19, 2009 (74 FR
23444). The supplemental letters dated September 21 and December 22,
2009, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 2010.
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 application for amendment: February 25, 2009.
Brief description of amendment: The changes remove the provisions
contained in Technical Specification (TS) 3/4.4.8, which specify
requirements relating to the structural integrity of American Society
of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components. This
specification is redundant to the requirements contained within Title
10 of the Code of Federal Regulations (10 CFR) Section 50.55a, ``Codes
and standards.'' With this change, the pressure boundary structural
integrity of ASME Code Class 1, 2 and 3
[[Page 10833]]
components will continue to be maintained through the facility's
compliance with 10 CFR 50.55a.
Date of issuance: February 24, 2010.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment Nos.: 199 and 160.
Facility Operating License Nos. NPF-39 and NPF-85. These amendments
revised the license and the technical specifications.
Date of initial notice in Federal Register: April 21, 2009 (74 FR
18254).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 24, 2010.
No significant hazards consideration comments received: No.
National Aeronautics and Space Administration, Docket Nos. 50-30, and
50-185. Erie County, Ohio
Date of amendment request: January 9, 2009, as supplemented by
letter dated October 6, 2009.
Brief description of amendment: The amendment adds a condition to
each license requiring that the National Aeronautics and Space
Administration assess the residual radioactivity and demonstrate that
the stream bed and banks of Plum Brook between the Plum Brook Station
boundary and Sandusky Bay meet the radiological criteria for
unrestricted use specified in 10 CFR 20.1402 prior to terminating
Licenses TR-3 and R-93.
Date of issuance: February 1, 2010.
Effective date: February 1, 2010.
Amendment Nos.: 14 and 10, respectively.
Possession Only License Nos. TR-3 and R-93: The amendment revises
both licenses.
Date of initial notice in Federal Register: May 5, 2009 (74 FR
20751).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation Report, dated February 1, 2010.
No Significant Hazards Consideration Comments Received: No.
National Aeronautics and Space Administration, Docket Nos. 50-30, and
50-185. Erie County, Ohio (TAC NO. J00301)
Date of amendment request: January 9, 2009, as supplemented by
letter dated October 6, 2009.
Brief description of amendment: The amendment adds a condition to
each license requiring that the National Aeronautics and Space
Administration assess the residual radioactivity and demonstrate that
the stream bed and banks of Plum Brook between the Plum Brook Station
boundary and Sandusky Bay meet the radiological criteria for
unrestricted use specified in 10 CFR 20.1402 prior to terminating
Licenses TR-3 and R-93.
Date of issuance: February 1, 2010.
Effective date: February 1, 2010.
Amendment Nos.: 14 and 10, respectively
Possession Only License Nos. TR-3 and R-93: The amendment revises
both licenses.
Date of initial notice in Federal Register: May 5, 2009 (74 FR
20751)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation Report, dated February 1, 2010.
No Significant Hazards Consideration Comments Received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: April 9, 2009.
Brief description of amendments: The amendments relocate Technical
Specification (TS) requirements pertaining to communications during
refueling operations (TS 3/4.9.5), manipulator crane operability (TS 3/
4.9.6), and crane travel (TS 3/4.9.7) to the Technical Requirements
Manual.
Date of issuance: February 17, 2010.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 293 and 277.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs and the License.
Date of initial notice in Federal Register: August 25, 2009 (74 FR
42929).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 17, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 25th day of February 2010.
For the Nuclear Regulatory Commission.
Allen G. Howe,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2010-4523 Filed 3-8-10; 8:45 am]
BILLING CODE 7590-01-P