[Federal Register Volume 75, Number 85 (Tuesday, May 4, 2010)]
[Notices]
[Pages 23808-23819]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-10105]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0169]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. 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 April 8, 2010 to April 21, 2010. The last
biweekly notice was published on April 20, 2010 (75 FR 20627).
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.
[[Page 23809]]
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, Rules,
Announcements and Directives Branch (RADB), 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 RADB 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 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
[[Page 23810]]
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
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].
Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: November 23, 2009.
Description of amendments request: The amendment would modify the
licensing basis and the Technical Specifications by allowing for the
transition from Westinghouse Turbo fuel to AREVA Advanced CE-14 High
Thermal Performance (HTP) fuel in the Calvert Cliffs reactors. The
licensee plans to refuel and operate with AREVA fuel beginning with the
refueling outage in 2011 for Unit No. 2 and 2012 for Unit No. 1. The
transition is planned to occur over three refueling cycles on each
unit.
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 23811]]
No.
The reactor fuel and the analyses associated with it are not
accident initiators. The response of the fuel to an accident is
analyzed using conservative techniques and the results are compared
to approved acceptance criteria. These evaluation results will show
that the fuel response to an accident is within approved acceptance
criteria for both cores loaded with the new AREVA Advanced CE-14 HTP
fuel and cores loaded with both AREVA and Westinghouse Turbo fuel.
Therefore, the change in fuel design does not affect accident or
transient initiation or consequences.
The proposed change to the Safety Limit Technical Specification
(2.1.1.2) does not require any physical change to any plant system,
structure, or component. The change to establish the peak fuel
centerline temperature as the safety limit is consistent with the
Standard Review Plan (SRP) for ensuring that the fuel design limits
are met. Operations and analysis will continue to be in compliance
with Nuclear Regulatory Commission (NRC) regulations. The peak fuel
centerline temperature is the basis for protecting the fuel and is
consistent with the analogous wording for other pressurized water
reactor (PWR) plants. Providing the peak fuel centerline melt
temperature as the safety limit does not impact the initiation or
the mitigation of an accident.
The proposed change to remove the total planar radial peaking
factor (F\T\XY, Technical Specification 3.2.2) is based
on a methodology change. During and after the transition to AREVA
Advanced CE-14 HTP fuel, the core analyses are performed using AREVA
methodologies. These methodologies do not use the total planar
radial peaking factor (F\T\XY) as an initial value in the
accident analyses. The linear heat rate algorithm limits are
provided by the total integrated radial peaking factor, azimuthal
power tilt, and axial shape index. The linear heat rate is evaluated
in accordance with NRC-approved methodology and meets acceptance
criteria. The total planar radial peaking factor is not an accident
initiator and does not play a role in accident mitigation. A number
of other changes are also made to remove references to Technical
Specification 3.2.2 throughout the Technical Specifications.
Topical reports have been reviewed and approved by the NRC for
use in determining core operating limits. The core operating limits
to be developed using the new methodologies will be established in
accordance with the applicable limitations as documented in the
appropriate NRC Safety Evaluation reports. The proposed change to
add and remove various topical reports to Technical Specification
5.6.5 enables the use of appropriate methodologies to re-analyze
certain events. The proposed methodologies will ensure that the
plant continues to meet applicable design criteria and safety
analysis acceptance criteria.
The proposed change to the list of NRC-approved methodologies
listed in Technical Specification 5.6.5 is administrative in nature
and has no impact on any plant configuration or system performance
relied upon to mitigate the consequences of an accident. The
proposed change will update the listing of NRC-approved
methodologies to remove methods no longer used and add new methods
consistent with the transition to AREVA Advanced CE-14 HTP fuel.
Changes to the calculated core operating limits may only be made
using NRC-approved methods, must be consistent with all applicable
safety analysis limits and are controlled by the 10 CFR 50.59
process. The list of methodologies in the Technical Specifications
does not impact either the initiation of an accident or the
mitigation of its consequences.
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 type of accident from any accident previously
evaluated?
No.
Use of AREVA Advanced CE-14 HTP fuel in the Calvert Cliffs
reactor cores is consistent with the current plant design bases and
does not adversely affect any fission product barrier, nor does it
alter the safety function of safety systems, structures, or
components, or their roles in accident prevention or mitigation. The
operational characteristics of AREVA Advanced CE-14 HTP fuel are
bounded by the safety analyses. The AREVA Advanced CE-14 HTP fuel
design performs within fuel design limits and does not create the
possibility of a new or different type of accident.
The proposed change to the Safety Limit Technical Specification
(2.1.1.2) does not require any physical change to any plant system,
structure, or component, nor does it require any change in safety
analysis methods or results. The existing analyses remain unchanged
and do not affect any accident initiators that would create a new
accident.
The proposed change to remove the total planar radial peaking
factor (F\T\XY, Technical Specification 3.2.2) is based
on a change in analytical methods needed to support the physical
fuel change. These methodologies do not use the total planar radial
peaking factor (F\T\XY) as an initial value in the
accident analysis. The total planar radial peaking factor does not
play a role in accident mitigation and cannot create the possibility
of a new or different kind of accident. A number of other changes
are made to remove references to Technical Specification 3.2.2
throughout the Technical Specifications.
The proposed change to the list of topical reports used to
determine the core operating limits is administrative in nature and
has no impact on any plant configuration or on system performance.
It updates the list of NRC-approved topical reports used to develop
the core operating limits. There is no change to the parameters
within which the plant is normally operated. The possibility of a
new or different accident is not created.
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?
No.
Use of AREVA Advanced CE-14 HTP fuel is consistent with the
current plant design bases and does not adversely affect any fission
product barrier, nor does it alter the safety function of safety
systems, structures, or components, or their roles in accident
prevention or mitigation. The operational characteristics of AREVA
Advanced CE-14 HTP fuel are bounded by the safety analyses. The
AREVA Advanced CE-14 HTP fuel design performs within fuel design
limits. The proposed changes do not result in exceeding design basis
limits. Therefore, all licensed safety margins are maintained.
The proposed change to the Safety Limit Technical Specification
(2.1.1.2) does not require any physical change to any plant system,
structure, or component, nor does it require any change in safety
analysis methods or results. Therefore, by changing the safety limit
from peak linear heat rate to peak fuel centerline temperature, the
margin as established in the current licensing basis remains
unchanged.
The proposed change to remove the total planar radial peaking
factor (F\T\XY,Technical Specification 3.2.2) is based on
a methodology change. The linear heat rate algorithm limits are
provided by the total integrated radial peaking factor, azimuthal
power tilt, and axial shape index. The linear heat rate is evaluated
in accordance with NRC-approved methodology and meets acceptance
criteria. Therefore, the margin as established for the linear heat
rate remains unchanged. A number of other changes are made to remove
references to Technical Specification 3.2.2 throughout the Technical
Specifications.
The proposed change to the list of topical reports does not
amend the cycle specific parameters presently required by the
Technical Specifications. The individual Technical Specifications
continue to require operation of the plant within the bounds of the
limits specified in the COLR [Core Operating Limits Report]. The
proposed change to the list of analytical methods referenced in the
COLR is administrative in nature and does not impact the margin of
safety.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Nancy L. Salgado.
[[Page 23812]]
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: March 15, 2010.
Description of amendment request: The proposed amendment would
revise a Technical Specification (TS) to address the increased
setpoints and setpoint tolerances for Safety Relief Valves (SRVs) and
Spring Safety Valves (SSVs) and changes related to the replacement of
four Target Rock two-stage SRVs with more reliable three-stage SRVs and
two existing Dresser 3.749 inch throat diameter SSVs with Dresser 4.956
inch diameter SSVs.
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 increases the allowable as-found SRV and SSV
setpoint tolerance, determined by test after the valves have been
removed from service, from 1% to 3%. The
proposed change also increases the SRV and SSV setpoints. Analysis
of these changes demonstrates that reactor pressure will be
maintained below the applicable code overpressure limits. The
proposed change increases the SSV discharge capacity due to its
increased throat diameter. The proposed change does not alter the TS
requirements for the number of SRVs and SSVs required to be
operable, the allowable as-left lift setpoint tolerance, the testing
frequency, or the manner in which the valves are operated.
Consistent with current TS requirements, the proposed change
continues to require that the safety valves be adjusted to within
1% of their nominal lift setpoints following testing.
The proposed increase in the SRV and SSV setpoint complies with the
ASME Boiler and Pressure Vessel (B&PV) Code (1965 Edition, including
January 1966 Addendum) for the pressure vessel, USAS Piping Code
Section B31.1 for the steam space piping, and ASME Section III for
the reactor coolant system recirculation piping. Since the proposed
change does not alter the manner in which the valves are operated,
there is no significant impact on the reactor operation.
The proposed change does not involve a change to the safety
function of the valves. The proposed TS revision involves no
significant changes to the operation of any systems or components in
normal or accident operating conditions. Therefore, these changes
will not increase the probability of an accident previously
evaluated.
Since an SSV setpoint increase and setpoint tolerance will
increase the SSV safety valve opening pressure and an increase in
the SSV throat size will increase the SSV flow capacity, the SSV
dynamic loads are expected to increase. Entergy has evaluated the
SSV dynamic loads for the associated piping. All piping and
structures were found to meet Code requirements.
Since an SRV setpoint and the setpoint tolerance increase will
increase the SRV valve opening pressure, the SRV discharge dynamic
loads will increase. Entergy has evaluated the SRV dynamic load
increases for the associated piping and torus submerged structures
and the evaluation concluded that all piping and structures were
found to meet Code requirements.
The proposed revision to the HPCI [high-pressure coolant
injection] and RCIC [Reactor Core Isolation Cooling] pump
operability determination surveillance follows the format of BWR
Standard Technical Specification surveillance, and complies with in-
service testing for pump operability determination in accordance
with ASME OM Code requirement.
Generic considerations related to the change in setpoints and
setpoint tolerance were addressed in NEDC-31753P, ``BWROG In-Service
Pressure Relief Technical Specification Revision Licensing Topical
Report,'' and were reviewed and approved by the NRC in a safety
evaluation dated March 8, 1993. General Electric Hitachi Company
(GEH) completed plant-specific analyses to assess the impact of
increase in SRV and SSV setpoints and increase in the setpoint
tolerance from 1% to 3%. The impact of the
increases in the SRV and SSV setpoints and increases in the setpoint
tolerances, as addressed in this analysis, included vessel
overpressure, Updated Final Safety Analysis Report (UFSAR) Chapter
14 events, ATWS [Anticipated Transient Without Scram], Loss of
Coolant Accident (LOCA), containment response and dynamic loads,
high-pressure systems performance, operating mode and equipment out
of service. The proposed change is supported by GEH analysis of
events that credit the SRVs and SSVs.
The plant specific evaluations, required by the NRC's safety
evaluation and performed to support this proposed change,
demonstrate that there is no change to the design core thermal
limits and adequate margin to the reactor coolant system pressure
limits exists. These analyses also demonstrate that operation of
Core Standby Cooling Systems (CSCS) is not adversely affected and
the containment response following a LOCA is acceptable. The plant
systems associated with these proposed changes are capable of
meeting applicable design basis requirements and retain the
capability to mitigate the consequences of accidents described in
the UFSAR. Therefore, these changes do not involve an increase in
the consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change increases the allowable as-found lift
setpoint tolerance for the Pilgrim SRV and SSV valves. The proposed
change to increase the tolerance was developed in accordance with
the provisions contained in the NRC safety evaluation for NEDC-
31753P. SRVs and SSVs installed in the plant following testing will
continue to meet the current tolerance acceptance criteria of 1% of the nominal setpoint. The proposed change does not
affect the manner in which the overpressure protection system is
operated; therefore, there are no new failure mechanisms for the
overpressure protection system.
The proposed changes do not change the safety function of the
SRVs and SSVs, or HPCI and RCIC systems. There is no alteration to
the parameters within which the plant is normally operated. The
increase in SRV and SSV setpoints, setpoint tolerance, and increased
SSV discharge capacity are not precursors to new or different kinds
of accidents and do not initiate new or different kinds of
accidents. The impact of these changes have been analyzed and found
to be acceptable within the design limits and plant operating
procedures.
As a result, 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 previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated and the establishment of the setpoints
for the actuation of equipment relied upon to respond to an event.
The proposed change modifies the setpoints at which protective
actions are initiated, and [* * *] does not change the requirements
governing operation or availability of safety equipment assumed to
operate to preserve the margin of safety.
Establishment of the 3% SRV and SSV setpoint
tolerance limit does not adversely affect the operation of any
safety-related component or equipment. Evaluations performed in
accordance with the NRC safety evaluation for NEDC-31753P have
concluded that all design limits will continue to be met.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy Salgado.
[[Page 23813]]
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 22, 2010.
Description of amendment request: The proposed amendment will
modify Technical Specification (TS) 3/4.9.4, ``Containment Building
Penetrations,'' to allow alternative means of penetration closure
during Core Alterations or irradiated fuel movement while in refueling
operations. Additional improvements to the TS are also being proposed,
as well as the elimination of TS 3/4.9.9, ``Containment Purge Valve
Isolation System.'' The proposed changes are consistent with Revision 3
of NUREG-1432, ``Standard Technical Specifications Combustion
Engineering Plants.''
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.
TS 3/4.9.4 currently allows containment penetration flow paths
to be open during Core Alterations or movement of irradiated fuel
within containment under specific administrative controls. The
proposed change would allow additional approved methods for ensuring
positive penetration closure. The fuel handling accident (FHA)
radiological analysis does not take credit for containment isolation
or filtration. Therefore, the time required to close any open
penetrations does not affect the radiological analysis dose
calculations and the proposed change does not involve a significant
increase in the consequences of an accident previously evaluated.
The administrative controls for containment penetration closure are
conservative even though not required by the accident analysis.
The proposed revision only provides alternate methods of
penetration closure and does not alter any plant equipment where the
probability of an accident would be increased. The incorporation of
purge valve isolation surveillance requirements for assuring purge
valve Operability has no effect on the probability or consequences
of the analyzed accidents.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Alternative methods of providing penetration closure do not
create accident initiators and do not represent a significant change
in the configuration of the plant. The proposed allowance to secure
containment penetrations during refueling operations will not
adversely effect plant safety functions or equipment operating
practices such that a new or different accident could be created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
TS Limiting Condition for Operation (LCO) 3.9.4 closure
requirements for containment penetrations ensure that the
consequences of a postulated FHA inside containment during Core
Alterations or fuel handling activities are minimized. The LCO
establishes containment closure requirements, which limit the
potential escape paths for fission products by ensuring that there
is at least one barrier to the release of radioactive material. The
proposed change to allow alternate methods of reaching containment
penetration closure during Core Alterations or fuel movement does
not affect the expected dose consequences of a FHA since it does not
credit containment building closure. The proposed administrative
controls provide assurance that prompt closure of the penetration
flow paths will be accomplished in the event of a FHA inside
containment thus minimizing the transmission of radioactive material
from the containment to the outside environment. The incorporation
of purge valve isolation surveillance requirements does not reduce
any margins 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: Joseph A. Aluise, Associate General
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 24, 2010.
Description of amendment request: The proposed amendment deletes
Operating License Condition 2.C.14 (Fuel Movement in the Fuel Handling
Building) due to electing to comply with Section 50.68, ``Criticality
accident requirements,'' of Title 10 of the Code of Federal Regulations
(10 CFR). The Operating License Condition 2.C.14, ``no more than one
fuel assembly shall be out of its shipping container or storage
location at a given time,'' was one basis for the exemption from the
criticality alarm system requirements of 10 CFR 70.24. The criticality
accident requirements can be met either by complying with 10 CFR 70.24
or 10 CFR 50.68 requirements. The 10 CFR 50.68 criteria are now being
used; therefore, Operating License Condition 2.C.14 is no longer
applicable.
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 amendment deletes Operating License Condition
2.C.14 (Fuel Movement in the Fuel Handling Building) due to electing
to comply with 10 CFR 50.68 requirements.
The proposed changes will not alter the configuration of the
storage racks or their environment. The fuel racks will not be
operated outside of their design limits, and no additional loads
will be imposed on them. Therefore, these changes will not affect
fuel storage rack performance or reliability. No new equipment will
be introduced into the plant. The accuracies and response
characteristics of existing instrumentation will not be modified.
The proposed changes will not require, or result in, a change in
safety system operation, and will not affect any system interface
with the fuel storage racks. Fuel assembly placement will continue
to be controlled in accordance with approved fuel handling
procedures. All the requirements of 10 CFR 50.68 continue to be met
which ensures no significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes will not affect any barrier that mitigates
dose to the public, and will not result in a new release pathway
being created. The functions of equipment designed to control the
release of radioactive material will not be impacted, and no
mitigating actions described or assumed for an accident in the UFSAR
[Updated Final Safety Analysis Report] will be altered or prevented.
No assumptions previously made in evaluating the consequences of an
accident will need to be modified. Onsite dose will not be
increased, so the access of plant personnel to vital areas of the
plant will not be restricted, and mitigating actions will not be
impeded.
Therefore, it is concluded that the proposed changes do not
significantly increase either the probability or consequences of any
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of
[[Page 23814]]
accident from any accident previously evaluated?
Response: No.
The proposed amendment deletes Operating License Condition
2.C.14 (Fuel Movement in the Fuel Handling Building) due to electing
to comply with 10 CFR 50.68 requirements.
10 CFR 50.68(b)(1) provides the requirements to ensure that
plant procedures shall prohibit the handling and storage at any one
time of more fuel assemblies than have been determined to be safely
subcritical under the most adverse moderation conditions feasible by
unborated water. By meeting this criteria, the removal of Operating
License Condition 2.C.14 will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, it is concluded that the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment deletes Operating License Condition
2.C.14 (Fuel Movement in the Fuel Handling Building) due to electing
to comply with 10 CFR 50.68 requirements.
10 CFR 50.68(b)(1) provides similar requirements as that
contained in Operating License Condition 2.C.14. The NRC has
approved the [Waterford Steam Electric Station, Unit 3] use of 10
CFR 50.68 criteria. By meeting the 10 CFR 50.68(b)(1) requirements,
there will not be a significant reduction in a margin of safety.
Therefore, it is 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, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: February 15, 2010.
Description of amendment request: The proposed amendment would
relocate selected Surveillance Requirement frequencies from the Clinton
Power Station, Unit No. 1 (Clinton) Technical Specifications (TSs) to a
licensee-controlled program. This change is based on the NRC-approved
Industry Technical Specifications Task Force (TSTF) change TSTF-425,
``Relocate Surveillance Frequencies to Licensee Control--Risk Informed
Technical Specification Task Force (RITSTF) Initiative 5b,'' Revision
3, (Agencywide Documents Access and Management System (ADAMS) Accession
Package No. ML090850642). Plant-specific deviations from TSTF-425 are
proposed to accommodate differences between the Clinton TSs and the
model TSs originally used to develop TSTF-425.
The Nuclear Regulatory Commission (NRC) staff issued a Notice of
Availability for TSTF-425 in the Federal Register on July 6, 2009 (74
FR 31996). The notice included a model safety evaluation (SE) and a
model no significant hazards consideration (NSHC) determination. In its
application dated February 15, 2010 (ADAMS Accession No. ML100470787),
the licensee affirmed the applicability of the model NSHC determination
which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No. The proposed change relocates the specified
frequencies for periodic surveillance requirements to licensee
control under a new Surveillance Frequency Control Program.
Surveillance frequencies are not an initiator to any accident
previously evaluated. As a result, the probability of any accident
previously evaluated is not significantly increased. The systems and
components required by the technical specifications for which the
surveillance frequencies are relocated are still required to be
operable, meet the acceptance criteria for the surveillance
requirements, and be capable of performing any mitigation function
assumed in the accident analysis. As a result, the consequences of
any accident previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No. No new or different accidents result from
utilizing the proposed change. The changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No. The design, operation, testing methods, and
acceptance criteria for systems, structures, and components (SSCs),
specified in applicable codes and standards (or alternatives
approved for use by the NRC) will continue to be met as described in
the plant licensing basis (including the final safety analysis
report and bases to TS), since these are not affected by changes to
the surveillance frequencies. Similarly, there is no impact to
safety analysis acceptance criteria as described in the plant
licensing basis. To evaluate a change in the relocated surveillance
frequency, Exelon will perform a probabilistic risk evaluation using
the guidance contained in NRC approved NEI 04-01, Rev. 1. The
methodology provides reasonable acceptance guidelines and methods
for evaluating the risk increase of proposed changes to surveillance
frequencies consistent with Regulatory Guide 1.177 [An Approach for
Plant-Specific, Risk-Informed Decision-making: Technical
Specifications].
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: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: March 3, 2010.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) 3.1.7, ``Standby Liquid Control (SLC)
System,'' to extend the completion time (CT) for Condition B (i.e.,
``Two SLC subsystems inoperable'') from 8 hours to 72 hours.
Basis for proposed no significant hazards consideration: As
required by 10 CFR 50.91(a), an analysis of the issue of no significant
hazards consideration is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or
[[Page 23815]]
consequences of any accident previously evaluated?
Response: No.
The proposed amendment revises Technical Specification (TS)
3.1.7, ``Standby Liquid Control (SLC) System,'' to extend the
completion time (CT) for Condition B (i.e., ``Two SLC subsystems
inoperable.'') from eight hours to 72 hours.
The proposed change is based on a risk-informed evaluation
performed in accordance with Regulatory Guides (RG) 1.174, ``An
Approach for Using Probabilistic Risk Assessment in Risk-Informed
Decisions On Plant-Specific Changes to the Licensing Basis,'' and RG
1.I77, ``An Approach for Plant-Specific, Risk-Informed Decision-
making: Technical Specifications.''
The proposed amendment modifies an existing CT for a dual-train
SLC system inoperability. The condition evaluated, the action
requirements, and the associated CT do not impact any initiating
conditions for any accident previously evaluated.
The proposed amendment does not increase postulated frequencies
or the analyzed consequences of an Anticipated Transient Without
Scram (ATWS). Requirements associated with 10 CFR 50.62 will
continue to be met. In addition, the proposed amendment does not
increase postulated frequencies or the analyzed consequences or a
large-break loss-of-coolant accident for which the SLC system will
be used for pH control. The extended CT provides additional time to
implement actions in response to a dual-train SLC system
inoperability, while also minimizing the risk associated with
continued operation. 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 previously evaluated?
Response: No.
The proposed amendment revises TS 3.1.7 to extend the CT for
Condition B from eight hours to 72 hours. The proposed amendment
does not involve any change to plant equipment or system design
functions. This proposed TS amendment does not change the design
function of the SLC system and does not affect the system's ability
to perform its design function. The SLC system provides a method to
bring the reactor, at any time in a fuel cycle, from full power and
minimum control rod inventory to a subcritical condition with the
reactor in the most reactive xenon free state without taking credit
for control rod movement. Required actions and surveillance
requirements are sufficient to ensure that the SLC system functions
are maintained. No new accident initiators are introduced by this
amendment. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment revises TS 3.1.7 to extend the CT for
Condition B from eight hours to 72 hours. The proposed amendment
does not involve any change to plant equipment or system design
functions. The margin of safety is established through the design of
the plant structures, systems, and components, the parameters within
which the plant is operated, and the setpoints for the actuation of
equipment relied upon to respond to an event.
The proposed amendment does not modify the condition or point at
which SLC is initiated, nor does it affect the system's ability to
perform its design function. In addition, the proposed change
complies with the intent of the defense-in-depth philosophy and the
principle that sufficient safety margins are maintained, consistent
with RG 1.177 requirements (i.e., Section C, ``Regulatory
Position,'' paragraph 2.2 ``Traditional Engineering
considerations'').
Based on the above analysis, EGC concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the analysis adopted by the licensee
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. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of amendment request: August 31, 2009.
Description of amendment request: The proposed amendment would
modify the PBAPS Technical Specifications (TS) by relocating specific
surveillance frequencies to a licensee-controlled program with the
implementation of Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed
Technical Specifications Initiative 5b, Risk-Informed Method for
Control of Surveillance Frequencies.'' Additionally, the change would
add a new program, the Surveillance Frequency Control Program, to TS
Section 5, Administrative Controls. The changes are based on NRC-
approved Industry Technical Specifications Task Force (TSTF) Traveler
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee
Control--Risk Informed Technical Specification Task Force Initiative
5b,'' with optional changes and variations as described in Attachment
1, Section 2.2 of the licensee's submittal dated August 31, 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. Do the proposed changes involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program [SFCP]. Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the technical specifications for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the surveillance requirements, and be
capable of performing any mitigation function assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
[* * * T]here is no impact to safety analysis acceptance
criteria as described in the plant licensing basis. To evaluate a
change in the relocated surveillance frequency, Exelon will perform
a probabilistic risk evaluation using the guidance contained in NRC
approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-
10, Rev. 1, methodology provides reasonable acceptance guidelines
and methods for evaluating the risk increase of proposed changes to
surveillance frequencies consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do
[[Page 23816]]
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, 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. J. Bradley Fewell, Associate General
Counsel, Exelon Generation Company LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
FPL Energy Seabrook, LLC Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 16, 2010.
Description of amendment request: The proposed changes would revise
the Seabrook Technical Specifications requirement that the Operations
Manager shall have held a senior reactor operator license for the
Seabrook Station prior to assuming the Operations Manager position.
Specifically, the proposed change would require the Operations Manager
to meet one of the following: (1) Hold a senior operator license; (2)
have held a senior operator license for a similar unit; or (3) have
been certified for equivalent senior operator knowledge. In its
application dated March 16, 2010, the licensee concluded that the no
significant hazards consideration (NSHC) determination presented in the
notice is applicable to Seabrook Station.
Basis for proposed NSHC determination: As required by 10 CFR
50.91(a), the licensee has provided its analysis of the issue of NSHC,
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 requested change would only affect the qualification
requirements for the Operations Manager Position]. The proposed
change does not impact the configuration or function of plant
structures, systems, or components (SSCs) or the manner in which
SSCs are operated, maintained, modified, tested, or inspected. No
actual facility equipment or accident analyses will be affected by
the proposed changes. Therefore, this request has no [significant]
impact on 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 accident previously
evaluated.
[The requested change would only affect the qualification
requirements for the Operations Manager Position]. The proposed
change does not alter the plant configuration, require new plant
equipment to be installed, alter accident analysis assumptions, add
any initiators, or affect the function of plant systems or the
manner in which systems are operated, maintained, modified, tested,
or inspected. Therefore, this request does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
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 requested change would
only affect the qualification requirements for the Operations
Manager Position]. No actual plant equipment or accident analyses
will be affected by the proposed changes. Additionally, the proposed
changes will not relax any criteria used to establish safety limits,
will not relax any safety system settings, and will not relax the
bases for any limiting conditions for operation. 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. The proposed change does not adversely affect
systems that respond to safely shutdown the plant and to maintain
the plant in a safe shutdown condition. Therefore, these 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, 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: M.S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Harold K. Chernoff.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP),
Goodhue County, Minnesota
Date of amendment request: November 24, 2009.
Description of amendment request: The proposed amendments would
make changes to Technical Specification (TS) Section 4.2.1, Fuel
Assemblies, and TS Section 5.6.5, Core Operating Limit Report, by
revising the TS to allow the use of Optimized ZIRLO\TM\ fuel rod
cladding material.
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.
Westinghouse Electric Company, LLC (Westinghouse) topical report
WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized
ZIRLO\TM\'', July 2006, provides the details and results of material
testing of Optimized ZIRLO\TM\ compared to standard ZIRLO\TM\ as
well as the material properties to be used in various models and
methodologies when analyzing Optimized ZIRLO\TM\. The Nuclear
Regulatory Commission (NRC) has allowed use of Optimized ZIRLO\TM\
fuel cladding material in Westinghouse fueled reactors provided that
licensees ensure compliance with the conditions and limitations set
forth in the NRC Safety Evaluation (SE) for the topical report. By
satisfying the conditions and limitations of the NRC SE through
completed actions and its approved reload safety evaluation process,
the licensee ensures that the effects of Optimized ZIRLO\TM\ on
PINGP core performance are evaluated and that the probability or
consequences of previously-evaluated accidents are not increased.
Therefore, the proposed change of adding a cladding material
does not result in an increase to 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.
Material properties of this fuel design have been evaluated in
Westinghouse topical report WCAP-12610-P-A and CENPD-404-P-A,
Addendum 1-A ``Optimized ZIRLO\TM\'' July 2006. That report provides
the details and results of material testing of Optimized ZIRLO\TM\
compared to standard ZIRLO\TM\ as well as the material properties to
be used in various models and methodologies when analyzing Optimized
ZIRLO\TM\. Neither that topical report nor the associated NRC SE
identifies the possibility of a new or different kind of accident
resulting from this change for generic application in Westinghouse
reactors. As demonstrated in that topical report and stated in the
NRC SE, there is reasonable assurance that under both normal and
accident conditions, the Optimized ZIRLO\TM\ fuel cladding will be
able to safely operate and comply with NRC regulations. By
satisfying the conditions and limitations of the NRC SE by virtue of
its completed actions and its approved reload safety evaluation
process, the licensee ensures that the effects of Optimized
ZIRLO\TM\ are evaluated and will not create the possibility of a new
or different kind of accident. Assurance that the possibility of new
or different type of accidents will not be created on a site-
specific basis is inherent to the reload safety evaluation process
approved for use at the PINGP. Site specific evaluation of the PINGP
core designs with Optimized ZIRLO\TM\ will be performed
programmatically and necessarily by the approved reload safety
evaluation process.
[[Page 23817]]
Therefore, the proposed change of adding a cladding material
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 cladding material used in the fuel rods is designed and
tested to prevent excessive fuel temperatures, excessive internal
rod gas pressure due to fission gas releases, and excessive cladding
stresses and strains. Optimized ZIRLO\TM\ was developed to meet
these needs and provides a reduced corrosion rate while maintaining
the benefits of mechanical strength and resistance to accelerated
corrosion from abnormal chemistry conditions. Westinghouse topical
report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized
ZIRLO\TM\, July 2006, provides the details and results of material
testing of Optimized ZIRLO\TM\ compared to standard ZIRLO\TM\ as
well as the material properties to be used in various models and
methodologies when analyzing Optimized ZIRLO\TM\. The NRC has
allowed use of Optimized ZIRLO\TM\ fuel cladding material detailed
within this topical report as detailed within their SE. Therefore,
the change in material does not result in a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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 J. Pascarelli.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: January 27, 2010.
Description of amendment request: The proposed amendments would
make changes to the Technical Specifications (TS) to revise TS 3.8.3,
``Diesel Fuel Oil''. The amendments would revise the diesel fuel oil
(DFO) storage volumes applicable to Unit 1 in TS 3.8.3 Condition
statements A and D, and increase the Unit 1 DFO supply required by
surveillance requirement 3.8.3.1. The amendments would clarify wording
in TS 3.8.3 Condition B statement which applies to both units.
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.
This license amendment request proposes to increase the
emergency diesel generator fuel oil storage volumes specified in the
Technical Specification Condition statements and Surveillance
Requirements. Also a word was added to a Condition statement to
clarify its meaning.
The emergency diesel generators and their supporting diesel fuel
oil storage systems are not accident initiators and therefore the
proposed fuel oil storage volume increases do not involve an
increase in the probability of an accident.
The proposed increased diesel fuel oil storage volumes provide
sufficient volumes to maintain the current licensing basis for
emergency diesel generator operation. Thus the proposed fuel oil
storage volume increases do not involve a significant increase in
the consequences of an accident.
The proposed Technical Specification Condition statement wording
clarification is administrative and thus does not involve an
increase in the probability of an accident or an increase in the
consequences of an accident.
Therefore, the proposed Technical Specification 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.
This license amendment request proposes to increase the
emergency diesel generator fuel oil storage volumes specified in the
Technical Specification Condition statements and Surveillance
Requirements. Also a word was added to a Condition statement to
clarify its meaning.
The proposed Technical Specification changes which increase
emergency diesel generator fuel oil storage volumes do not change
any system operations or maintenance activities. The changes do not
involve physical alteration of the plant, that is, no new or
different type of equipment will be installed. The changes do not
alter assumptions made in the safety analyses but ensures that the
diesel generators operate as assumed in the accident analyses. These
changes do not create new failure modes or mechanisms which are not
identifiable during testing and no new accident precursors are
generated.
The proposed Technical Specification Condition statement wording
clarification is administrative and thus does not create the
possibility of a new or different kind of accident.
Therefore, the proposed Technical Specification changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This license amendment request proposes to increase the
emergency diesel generator fuel oil storage volumes specified in the
Technical Specification Condition statements and Surveillance
Requirements. Also a word was added to a Condition statement to
clarify its meaning.
Since this license amendment proposes Technical Specification
changes which increase the required fuel oil storage volumes,
margins of safety are increased and thus no margin of safety is
reduced as part of this change.
The proposed Technical Specification Condition statement wording
clarification is administrative and thus does not involve a
significant reduction in a margin of safety.
Therefore, the proposed Technical Specification 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 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 J. Pascarelli.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: February 2, 2010.
Description of amendment request: The proposed amendments would
revise the verification requirements for the Reactor Trip System
Instrumentation. Specifically, the amendment proposes the addition to
Table 3.3.1-1 of a response time measurement for the verification of
the Power Range Neutron High Positive Rate Trip (PFRT) function as
recommended by Westinghouse Nuclear Safety Advisory Letter (NSAL-09-01)
``Rod Withdrawal at Power Analysis for Reactor Coolant System
Overpressure.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to Vogtle Electric Generating Plant (VEGP)
Technical Specification (TS) 3.3.1, ``Reactor Trip
[[Page 23818]]
System (RTS) Instrumentation,'' Table 3.3.1-1, ``Reactor Trip System
Instrumentation'' does not significantly increase the probability or
consequences of an accident previously evaluated in the Updated
Final Safety Analysis Report (UFSAR). The overall protection system
performance will remain within the bounds of the accident analysis
since there are no hardware changes. The design of the Reactor Trip
System (RTS) instrumentation, specifically the positive range
neutron flux high positive rate trip (PFRT) function, will be
unaffected. The reactor protection system will continue to function
in a manner consistent with the plant design basis. All design,
material, and construction standards that were applicable prior to
the request are maintained.
The proposed change adds an additional surveillance requirement
to assure that the PFRT is verified to be consistent with the safety
analysis and licensing basis. In this specific case, a response time
verification requirement will be added to the PFRT function.
The proposed changes will not modify any system interface. The
proposed changes will not affect the probability of any event
initiators. There will be no degradation in the performance of or an
increase in the number of challenges imposed on safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance. The proposed change will not alter any
assumptions nor change any mitigation actions in the radiological
consequences evaluations in the UFSAR.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes do not
alter nor prevent the ability of SSCs from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change is consistent
with the safety analyses assumptions and resultant consequences. The
RCS overpressure limit listed in Specification 2.1.2 of the VEGP
Technical Specifications (i.e., 2735 psig) is not violated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. This change will not affect the normal method of plant
operation nor change any operating parameters.
No performance requirements will be affected; however, the
proposed change adds an additional surveillance requirement. The
additional surveillance requirement is consistent with assumptions
made in the safety analyses and licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this change.
Therefore, the proposed change does not create the possibility
of a new or different 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 affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Limits. There
will be no effect on the manner in which Safety Limits or Limiting
Conditions of Operations are determined, nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions.
This change is consistent with the assumptions made in the
safety analyses. The addition of a surveillance requirement
increases the margin of safety by assuring that the associated
safety analysis assumption on the PFRT response time is verified.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based on the above, SNC concludes that the proposed amendment
does not involve a significant hazards consideration under the
standard set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of amendment request: March 31, 2010.
Brief description of amendment request: The proposed amendment
would add new license condition 2.C(4) stating that performance of
Technical Specification surveillance requirement 3.1.4.3, which
verifies control rod freedom of movement, is not required for control
rod drive 22 during cycle 21 until the next entry into Mode 3 in a
maintenance or refueling outage, whichever is earlier.
Date of publication of individual notice in Federal Register: April
14, 2010 (75 FR 19428).
Expiration date of individual notice: June 13, 2010.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: March 29, 2010, as supplemented by
letter dated March 29, 2010.
Brief description of amendment request: The proposed amendment
would revise the Technical Specification (TS) 3.3.2, ``Engineered
Safety Feature Actuation System (ESFAS) Instrumentation,'' regarding
function 6.g in TS Table 3.3.2-1. Function 6.g provides an auxiliary
feedwater (AFW) start signal that is provided to the motor-driven AFW
pumps in the event of a trip of both turbine-driven main feedwater
pumps. The changes would revise Condition J for ESFAS instrumentation
function 6.g to read, ``One or more Main Feedwater Pumps trip
channel(s) inoperable.'' The licensee will make corresponding changes
to Required Action J.1 and the Note above Required Actions J.1 and J.2
for consistency with the revised Condition.
Date of publication of individual notice in Federal Register: April
14, 2010 (75 FR 19431).
Expiration date of individual notice: April 28, 2010, for public
comments; June 14, 2010, for hearing requests.
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
[[Page 23819]]
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. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: November 23, 2009, as
supplemented by letter dated February 5, 2010.
Brief description of amendment: The amendment modified the
Technical Specification (TS) 5.5.7, ``Inservice Testing Program,'' by
replacing the references from the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code to the current Code of
Record, the ASME Operation and Maintenance Nuclear Power Plants Code
(ASME OM Code), the Code of Record for the James A. FitzPatrick Nuclear
Power Plant (JAFNPP) Inservice Testing (IST) Program. This is an
administrative amendment to maintain the TS current with the NRC
accepted Code of Record for JAFNPP IST Program.
Date of issuance: April 12, 2010.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 296.
Renewed Facility Operating License No. DPR-59: The amendment
revised the License and the Technical Specifications.
Date of initial notice in Federal Register: January 26, 2010 (75 FR
4117).
The February 5, 2010, supplement 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 April 12, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station (Byron), Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendment: September 24, 2009, as
supplemented by letters dated November 13, 2009; January 19, 2010;
March 1, 2010; March 9, 2010 (two letters); and March 19, 2010.
Brief description of amendment: The amendments adds a new
Completion Time (CT) of 144 hours to restore a unit-specific essential
service water train to operable status associated with the Limiting
Condition for Operation for Technical Specification (TS) 3.7.8,
``Essential Service Water (SX) System.'' The new CT will be used for
maintenance during the Byron, Unit No. 2, spring 2010, refueling
outage. The licensee requested the new CT to replace two of the four SX
pump suction isolation valves without having to shutdown Byron, Unit
No. 1; maintenance history has shown that replacement of the SX pump
suction isolation valves cannot be assured within the existing 72 hour
CT window.
Date of issuance: April 9, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: Unit No. 1--168; Unit No. 2--168.
Facility Operating License Nos. NPF-37 and NPF-66: The amendments
revise the TSs and Licenses.
Date of initial notice in Federal Register: December 1, 2009 (74 FR
62835).
The supplemental letters 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 9, 2010.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: September 18, 2009.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.5.7, ``Inservice Testing Program,'' by
incorporating TS Task Force Traveler (TSTF)-479, ``Changes to Reflect
Revision of 10 CFR 50.55a,'' and TSTF-497, ``Limit Inservice Testing
Program SR [Surveillance Requirement] 3.0.2 Application to Frequencies
of 2 Years or Less.'' Specifically, the amendments (1) replace
references to the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code, Section XI with the ASME Code for
Operation and Maintenance of Nuclear Power Plants for inservice testing
activities, and (2) applies the extension allowance of SR 3.0.2 to
other normal and accelerated inservice testing frequencies of 2 years
or less that were not included in the frequencies listed in TS 5.5.7.a.
Date of issuance: April 8, 2010.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 110.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: November 3, 2009 (74 FR
56887).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 22nd day of April 2010.
For the Nuclear Regulatory Commission.
Robert A. Nelson,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2010-10105 Filed 5-3-10; 8:45 am]
BILLING CODE 7590-01-P