[Federal Register Volume 74, Number 229 (Tuesday, December 1, 2009)]
[Notices]
[Pages 62831-62840]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-28630]
[[Page 62831]]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0518]
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 November 5, 2009, to November 18, 2009. The
last biweekly notice was published on November 17, 2009 (74 FR 59259).
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 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 petitioner/requestor
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
petitioner/requestor 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 petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor 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 petitioner/requestor to relief. A petitioner/
requestor 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
[[Page 62832]]
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, which the NRC
promulgated in August 28, 2007 (72 FR 49139). 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 petitioner/requestor
should contact the Office of the Secretary by e-mail at
[email protected], or by calling (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/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM is free and is
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. 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.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE 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
EIE 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 through the ``Contact Us'' link
located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC Meta-System Help Desk, which is
available between 8 a.m. and 8 p.m., Eastern Time, Monday through
Friday, excluding government holidays. The Meta-System Help Desk can be
contacted by telephone at 1-866-672-7640 or by e-mail at
[email protected].
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.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the request and/
or petition should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii).
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, an Atomic Safety and Licensing Board, or
a 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 submissions.
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].
[[Page 62833]]
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: September 28, 2009.
Description of amendment request: The amendments would revise
Required Action A.1 of Technical Specification (TS) 3.8.7,
``Inverters--Operating,'' for the Palo Verde Nuclear Generating Station
(PVNGS), Units 1, 2, and 3, by extending the Completion Time for
restoration of an inoperable vital alternating current (AC) inverter
from 24 hours to 7 days.
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 TS amendment does not affect the design of the
vital AC inverters, the operational characteristics or function of
the inverters, the interfaces between the inverters and other plant
systems, or the reliability of the inverters. An inoperable vital AC
inverter is not considered an initiator of an analyzed event. In
addition, Required Actions and the associated Completion Times are
not initiators of previously evaluated accidents. Extending the
Completion Time for an inoperable vital AC inverter would not have a
significant impact on the frequency of occurrence of an accident
previously evaluated. The proposed amendment will not result in
modifications to plant activities associated with inverter
maintenance, but rather, provides operational flexibility by
allowing additional time to perform inverter troubleshooting,
corrective maintenance, and post-maintenance testing on-line.
The proposed extension of the Completion Time for an inoperable
vital AC inverter will not significantly affect the capability of
the inverters to perform their safety function, which is to ensure
an uninterruptible supply of 120-volt AC electrical power to the
associated power distribution subsystems. An evaluation, using PRA
[probabilistic risk assessment] methods, confirmed that the increase
in plant risk associated with implementation of the proposed
Completion Time extension is consistent with the NRC's Safety Goal
Policy Statement, as further described in [NRC Regulatory Guide] RG
1.174 and RG 1.177. In addition, a deterministic evaluation
concluded that plant defense-in-depth philosophy will be maintained
with the proposed Completion Time extension. Based on the above, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not involve physical alteration of
the PVNGS. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
is no change being made to the parameters within which the PVNGS is
operated. There are no setpoints at which protective or mitigating
actions are initiated that are affected by this proposed action. The
use of the alternate Class 1E power source for the vital AC
instrument bus is consistent with the PVNGS plant design. The change
does not alter assumptions made in the safety analysis. This
proposed action will not alter the manner in which equipment
operation is initiated, nor will the functional demands on credited
equipment be changed. No alteration is proposed to the procedures
that ensure the PVNGS remains within analyzed limits, and no change
is being made to procedures relied upon to respond to an off-normal
event. As such, no new failure modes are being introduced.
Based on the above, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of setpoints to initiate alarms
or actions. The proposed amendment does not alter the design or
configuration of the vital AC inverters or their associated 120-volt
AC subsystems, and does not alter the setpoints at which alarms and
associated actions are initiated. With one of the required 120-volt
AC vital instrumentation buses being powered from the alternate
safety-related Class 1E power supply, which is backed by the
divisional diesel generator (DG), there is no significant reduction
in the margin of safety. Testing of the DGs and associated
electrical distribution equipment provides confidence that the DGs
will start and provide power to the associated equipment in the
unlikely event of a loss of offsite power during the extended 7-day
Completion Time.
Applicable regulatory requirements will continue to be met,
adequate defense-in-depth will be maintained, sufficient safety
margins will be maintained, and any increases in risk are consistent
with the NRC Safety Goal Policy Statement. Furthermore, during the
proposed extended inverter Completion Time, any increases in risk
posed by potential combinations of equipment out of service will be
managed in accordance with the PVNGS site Configuration Risk
Management Program, consistent with Paragraph (a)(4) of 10 CFR
50.65, ``Requirements for monitoring the effectiveness of
maintenance at nuclear power plants.''
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: August 18, 2009.
Description of amendments request: The proposed license amendments
revise Technical Specification 3.3.1.1, ``Reactor Protection System
(RPS) Instrumentation,'' Surveillance Requirement 3.3.1.1.8, to
increase the frequency interval between local power range monitor
calibrations from 1100 megawatt-days per metric ton average core
exposure (i.e., equivalent to approximately 907 effective full-power
hours (EFPH)) to 2000 EFPH.
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 amendments revise the surveillance interval for the
LPRM [local power range monitor] calibration from 1100 MWD/T
[megawatt days per metric ton] average core exposure to 2000
effective full power hours (EFPH). Increasing the frequency interval
between required LPRM calibrations is acceptable due to improvements
in fuel analytical bases, core monitoring processes, and nuclear
instrumentation. The revised surveillance interval continues to
ensure that the LPRM detector signal will continue to be adequately
calibrated.
This change will not alter the operation of process variables,
structures, systems, or components as described in the Updated Final
Safety Analysis Report. The probability of an evaluated accident is
derived from the probabilities of the individual precursors to that
accident. The proposed change does not alter the initiation
conditions or operational parameters for the LPRM subsystem and
there is no new equipment introduced by the
[[Page 62834]]
extension of the LPRM calibration interval. The performance of the
Average Power Range Monitor (APRM), Rod Block Monitor (RBM), and
Oscillation Power Range Monitor (OPRM) systems is not affected by
the proposed surveillance interval increase. The proposed LPRM
calibration interval extension will have no significant effect on
the Reactor Protection System (RPS) instrumentation accuracy during
power maneuvers or transients and will, therefore, not significantly
affect the performance of the RPS. As such, no individual precursors
of an accident are affected and the proposed amendments do not
increase the probability of a previously analyzed event.
The radiological consequences of an accident can be affected by
the thermal limits existing at the time of the postulated accident;
however, increasing the surveillance interval frequency will not
increase the calculated thermal limits since all uncertainties
associated with the increased interval are currently implemented and
are currently used to calculate the existing safety limits. Plant
specific evaluation of LPRM sensitivity to exposure has determined
that the extended calibration frequency increases the LPRM signal
uncertainty value used in the SLMCPR [safety limit for minimum
critical power] analysis; however, the increase is bounded by the
values currently used in the safety analysis. Therefore, the thermal
limit calculation is not significantly affected by LPRM calibration
frequency, and thus the radiological consequences of any accident
previously evaluated are not increased.
Based on the above, the proposed amendments do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident requires creating one or more new accident precursors. New
accident precursors may be created by modifications of plant
configuration, including changes in allowable modes of operation.
The performance of the APRM, RBM, and OPRM systems are not affected
by the proposed LPRM surveillance interval increase. The proposed
change does not affect the control parameters governing unit
operation or the response of plant equipment to transient
conditions. For the proposed LPRM extended calibration interval
frequency, all uncertainties remain less than the uncertainties
assumed in the existing thermal limit calculations. The proposed
change does not change or introduce any new equipment, modes of
system operation, or failure mechanisms; therefore, no new accident
precursors are created. Based on the above information, the proposed
amendments do not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change has no impact on equipment design or
fundamental operation, and there are no changes being made to safety
limits or safety system allowable values that would adversely affect
plant safety as a result of the proposed LPRM surveillance interval
increase. The performance of the APRM, RBM, and OPRM systems are not
affected by the proposed change. The margin of safety can be
affected by the thermal limits existing at the time of the
postulated accident; however, uncertainties associated with LPRM
chamber exposure have no significant effect on the calculated
thermal limits. Plant-specific evaluation of LPRM sensitivity to
exposure has determined that the extended calibration frequency
increases the LPRM signal uncertainty value used in the SLMCPR
analysis; however, the increase is bounded by the values currently
used in the safety analysis. The thermal limit calculation is not
significantly affected since LPRM sensitivity with exposure is well
defined. LPRM accuracy remains within that used to determine the
total power uncertainty assumed in the thermal analysis basis,
therefore maintaining thermal limits and the safety margin. The
proposed change does not affect uncertainties or initial conditions
assumed in the thermal limit calculations and therefore the margin
of safety in the safety analyses is maintained. Based on the above
information, the proposed amendments do not result in a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 19, 2009.
Description of amendment request: The proposed amendment relocates
the Waterford Steam Electric Station, Unit 3 Steam Generator Level--
High trip requirements from Technical Specification Sections 2.2 and 3/
4.3.1 to the Technical Requirements Manual.
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 relocates the Steam Generator Level--High
Trip to a licensee-controlled document. The Steam Generator (SG)
Level--High trip function is not credited in any DBA [design-basis
accident] or transient analysis and is not an initiator to any
accident analysis. As a result, neither the probability nor the
consequences of an accident previously evaluated are significantly
increased by this change.
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.
The proposed change relocates the Steam Generator Level--High
trip function to a licensee-controlled document. The proposed change
does not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change relocates the Steam Generator Level--High
trip function to a licensee-controlled document. This will allow
changes to the Steam Generator Level--High Trip requirements
currently in the Technical Specifications to be performed in
accordance with the requirements of 10 CFR 50.59. As the Steam
Generator Level--High trip function has been determined to not meet
the definition of Technical Specifications or the criteria in 10 CFR
50.36 (c)(2)(ii), lack of NRC review and approval prior to
implementation for changes that are not determined to be a
significant hazard will not lead to a significant reduction in the
margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
[[Page 62835]]
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: September 24, 2009.
Description of amendment request: The amendment request proposes a
one-time extension of the Completion Time (CT) to restore a unit-
specific essential service water train to operable status associated
with Technical Specification Limiting Condition for Operation (LCO)
3.7.8, Essential Service Water (SX) System, from 72 hours to 144 hours.
The proposed change will only be used one time during the Byron Station
Unit 2 spring 2010 refueling outage. The licensee is requesting an
extension of the CT to 144 hours to replace two of the four SX pump
suction isolation valves; maintenance history has shown that
replacement of the SX pump suction isolation valves cannot be assured
within the existing 72 hour CT window.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes have been evaluated using the risk-informed
processes described in Regulatory Guide (RG) 1.174, ``An Approach
for Using Probabilistic Risk Assessment in Risk-Informed Decisions
on Plant-Specific Changes to the Licensing Basis,'' dated July 1998
and RG 1.177, ``An Approach for Plant-Specific, Risk-Informed
Decisionmaking: Technical Specifications,'' dated August 1998. In
addition, proposed revised guidance as described in Draft Regulatory
Guide DG-1226, ``An Approach for Using Probabilistic Risk Assessment
in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis,'' and Draft Regulatory Guide DG-1227, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' was reviewed for insights. The risk associated
with the proposed changes was shown to be acceptable.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The SX system is not
considered an initiator for any of these previously analyzed events.
The proposed change does not have a detrimental impact on the
integrity of any plant structure, system, or component that
initiates an analyzed event. No active or passive failure mechanisms
that could lead to an accident are affected. The proposed change
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident. Therefore, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
The unit-specific SX system consists of two separate,
electrically independent, 100% capacity, safety related, cooling
water trains. Each train consists of a 100% capacity pump, piping,
valving, and instrumentation. Normally, the pumps and valves are
remotely and manually aligned. However, the pumps are automatically
started upon receipt of a safety injection signal or an undervoltage
on the engineered safety features (ESF) bus, and all essential
valves are aligned to their post accident positions. The SX system
is also the backup water supply to the auxiliary feedwater system
and fire protection system.
The design basis of the SX system is for one SX train, in
conjunction with the component cooling water (CC) system and a 100%
capacity containment cooling system, to remove core decay heat
following a design basis LOCA [loss-of-coolant accident] as
discussed in the UFSAR [updated final safety analysis report],
Section 6.2, ``Containment Systems.'' This prevents the containment
sump fluid from increasing in temperature during the recirculation
phase following a LOCA and provides for a gradual reduction in the
temperature of this fluid as it is supplied to the reactor coolant
system by the emergency core cooling system pumps. The SX system is
designed to perform its function with a single failure of any active
component, assuming the loss of offsite power. The proposed one-time
increase in the CT is consistent with the philosophy of the current
Technical Specification LCO which allows one train of SX to be
inoperable for 72 hours. This change only extends the 72 hour
Completion Time to 144 hours which has been shown to be acceptable
from a risk perspective; therefore, the proposed change does not
involve a significant increase in the 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 changes do not involve the use or installation of
new equipment and the currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bases. Based on this evaluation, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter any existing setpoints at
which protective actions are initiated and no new setpoints or
protective actions are introduced. The design and operation of the
SX system remains unchanged. The risk associated with the proposed
increase in the time an SX pump is allowed to be inoperable was
evaluated using the risk-informed processes described in RG 1.174,
``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing
Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications,''
dated August 1998. The risk was shown to be acceptable. Based on
this evaluation, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (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.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334,
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County,
Pennsylvania
Date of amendment request: July 6, 2009.
Description of amendment request: The proposed amendment would
revise Technical Specification 5.6.3, ``Core Operating Limits Report,''
to allow the use of the generically approved Topical Report, WCAP-
16009-P-A, ``Realistic Large Break LOCA [Loss-of-Coolant Accident]
Evaluation Methodology Using Automated Statistical Treatment of
Uncertainty Method (ASTRUM),'' for BVPS-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. No physical changes are required as a result of implementing
the ASTRUM best-estimate large break [LOCA] methodology and
associated technical specification changes. The plant conditions
assumed in the analysis are bounded by the design conditions for all
equipment in the plant. Therefore, there will be no increase in the
probability of a LOCA. The consequences of a LOCA are not being
increased, since it is shown that the emergency core cooling system
is designed so that its calculated cooling performance conforms to
the criteria contained in 10 CFR 50.46, Paragraph (b). No
[[Page 62836]]
other accident is potentially affected by this change.
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?
No. There are no physical changes being made to the plant. No
new modes of plant operation are being introduced. The parameters
assumed in the analysis are within the design limits of the existing
plant equipment. All plant systems will perform as designed during
the response to a potential accident.
Therefore, the proposed change does not involve an increase in
the probability or consequences of an accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The methodology used in the analysis would more
realistically describe the expected behavior of plant systems during
a postulated loss of coolant accident. Uncertainties have been
accounted for as required by 10 CFR 50.46. A sufficient number of
loss of coolant accidents with different break sizes, different
locations and other variations in properties are analyzed to provide
assurance that the most severe postulated LOCAs are calculated. As
described in Section 3.3, there is a high level of probability that
all criteria contained in 10 CFR 50.46, Paragraph (b) are met.
The Nuclear Regulatory Commission (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, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: Nancy L. Salgado.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: September 9, 2009.
Description of amendment request: The proposed amendment would
change the frequency of control rod notch testing, as specified in
Technical Specification (TS) surveillance requirement (SR) 4.1.3.1.2.a,
from at least once per 7 days to at least once per 31 days. The purpose
of this SR is to confirm control rod insertion capability which is
demonstrated by inserting each partially or fully withdrawn control rod
at least one notch and observing that the control rod moves. This
ensures that the control rod is not stuck and is free to insert on a
scram signal. The proposed amendment would also add the word ``fully''
to the Action for TS Limiting Condition for Operation (LCO) 3.9.2 to
clarify the requirement to fully insert all insertable control rods
when the required source range monitor (SRM) instrumentation is
inoperable. The licensee stated that the proposed amendment is based on
Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF)
change, TSTF-475, Revision 1, ``Control Rod Notch Testing Frequency and
SRM Insert Control Rod Action.'' The availability of this change to the
Standard Technical Specifications (STS) was announced in the Federal
Register on November 13, 2007 (72 FR 63935) as part of the consolidated
line item improvement process. The Federal Register notice included a
model safety evaluation, a model application and a model proposed a no
significant hazards consideration (NSHC) determination. In its
application dated September 9, 2009, the licensee affirmed the
applicability of the proposed NSHC determination for TSTF-475 and has
incorporated it by reference to satisfy the requirements of 10 CFR
50.91(a). Since Hope Creek Generating Station has not adopted the STS
(e.g., NUREG-1433), the licensee has proposed minor variations from the
TS changes described in TSTF-475.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff's review 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 SR 4.1.3.1.2.a reduces the frequency of
control rod notch testing. Changing the frequency of testing is not
expected to have any significant impact on the reliability of the
control rods to insert as required on a scram signal. The proposed
change to the Action for LCO 3.9.2 merely clarifies the intent of the
action. There are no physical plant modifications associated with this
change. The proposed amendment would not alter the way any structure,
system, or component (SSC) functions and would not alter the way the
plant is operated. As such, the proposed amendment would have no impact
on the ability of the affected SSCs to either preclude or mitigate 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 proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment would not change the design function or
operation of the SSCs involved and would not impact the way the plant
is operated. As such, the proposed change would not introduce any new
failure mechanisms, malfunctions, or accident initiators not already
considered in the design and licensing bases. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is associated with the confidence in the
ability of the fission product barriers (i.e., fuel cladding, reactor
coolant pressure boundary, and containment structure) to limit the
level of radiation to the public. There are no physical plant
modifications associated with the proposed amendment. The proposed
amendment would not alter the way any SSC functions and would not alter
the way the plant is operated. The proposed amendment would not
introduce any new uncertainties or change any existing uncertainties
associated with any safety limit. The proposed amendment would have no
impact on the structural integrity of the fuel cladding, reactor
coolant pressure boundary, or containment structure. Based on the above
considerations, the NRC staff concludes that the proposed amendment
would not degrade the confidence in the ability of the fission product
barriers to limit the level of radiation to the public. Therefore, the
proposed change does not involve a significant reduction in a margin of
safety.
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: Vincent Zabielski, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: October 20, 2009.
[[Page 62837]]
Description of amendment request: The proposed amendment would
delete paragraph d of Technical Specification 5.2.2, ``Unit Staff,''
superseded by Title 10 of the Code of Federal Regulations Part 26,
Subpart I.
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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change removes Technical Specification (TS)
restrictions on working hours for personnel who perform safety
related functions. The TS restrictions are superseded by the worker
fatigue requirements in 10 CFR Part 26. The proposed change does not
impact the physical configuration or function of plant structures,
systems, or components (SSCs) or the manner in which SSCs are
operated, maintained, modified, tested, or inspected. Worker fatigue
is not an initiator of any accident previously evaluated. Worker
fatigue is not an assumption in the consequence mitigation of any
accident previously evaluated.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change removes TS restrictions on working hours for
personnel who perform safety related functions. The TS restrictions
are superseded by the worker fatigue requirements in 10 CFR Part 26.
Working hours will continue to be controlled in accordance with NRC
requirements. The new rule allows for deviations from controls to
mitigate or prevent a condition adverse to safety or as necessary to
maintain the security of the facility. This ensures that the new
rule will not unnecessarily restrict working hours and thereby
create the possibility of a new or different kind of accident from
any accident previously evaluated.
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, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change removes TS restrictions on working hours for
personnel who perform safety related functions. The TS restrictions
are superseded by the worker fatigue requirements in 10 CFR Part 26.
The proposed change does not involve any physical changes to plant
or alter the manner in which plant systems are operated, maintained,
modified, tested, or inspected. 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. The proposed change does not
adversely affect systems that respond to safely shut down the plant
and to maintain the plant in a safe shut down condition.
Removal of plant-specific TS administrative requirements will
not reduce a margin of safety because the requirements in 10 CFR
Part 26 are adequate to ensure that worker fatigue is managed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas Boyce.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 20, 2009.
Description of amendment request: The proposed amendment would
delete paragraph g of Technical Specification 6.2.2, ``Facility
Staff,'' which was superseded by Title 10 of the Code of Federal
Regulations (10 CFR), Part 26, Subpart I. This change is consistent
with Nuclear Regulatory Commission approved Technical Specification
Task Force (TSTF) Improved Standard Technical Specification Change
Traveler TSTF-511, Revision 0, ``Eliminate Working Hour Restrictions
from TS 5.2.2 to Support Compliance with 10 CFR Part 26.''
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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change removes Technical Specification (TS)
restrictions on working hours for personnel who perform safety
related functions. The TS restrictions are superseded by the worker
fatigue requirements in 10 CFR Part 26. The proposed change does not
impact the physical configuration or function of plant structures,
systems, or components (SSCs) or the manner in which SSCs are
operated, maintained, modified, tested, or inspected. Worker fatigue
is not an initiator of any accident previously evaluated. Worker
fatigue is not an assumption in the consequence mitigation of any
accident previously evaluated.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change removes TS restrictions on working hours for
personnel who perform safety related functions. The TS restrictions
are superseded by the worker fatigue requirements in 10 CFR Part 26.
Working hours will continue to be controlled in accordance with NRC
requirements. The new rule allows for deviations from controls to
mitigate or prevent a condition adverse to safety or as necessary to
maintain the security of the facility. This ensures that the new
rule will not unnecessarily restrict working hours and thereby
create the possibility of a new or different kind of accident from
any accident previously evaluated.
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, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change removes TS restrictions on working hours for
personnel who perform safety related functions. The TS restrictions
are superseded by the worker fatigue requirements in 10 CFR Part 26.
The proposed change does not involve any physical changes to plant
or alter the manner in which plant systems are operated, maintained,
modified, tested, or inspected. 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. The proposed change does not
adversely affect systems that respond to safely shut down the plant
and to maintain the plant in a safe shutdown condition.
Removal of plant specific TS administrative requirements will
not reduce a margin of safety because the requirements
[[Page 62838]]
in 10 CFR Part 26 are adequate to ensure that worker fatigue is
managed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia
Date of amendment request: September 28, 2009.
Description of amendment request: The proposed changes would
address the filtration function of the Emergency Core Cooling System
(ECCS) Pump Room Exhaust Air Cleanup System (PREACS) and are consistent
with the associated design and licensing basis accident analysis
assumptions. The proposed changes will add new Conditions B and C with
associated Action Statements and Completion Times to Technical
Specification (TS) 3.7.12 and modify Conditions A and D.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed changes do not adversely affect accident initiators
or precursors and do not alter the design assumptions, conditions,
or configuration of the facility. The new conditions only affect the
filtration function of ECCS PREACS, which is an accident mitigation
function, so accident initiation probability is not impacted.
Regarding significance of the proposed changes relative to the
accident consequences, the new conditions remain consistent with
existing design assumptions (i.e., dose calculations show that the
filtration function is not required when ECCS leakage is less than
the maximum allowable unfiltered leakage) and filtration is required
to be operable as required to support the design analysis
assumptions.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The addition of the new Conditions B and C with associated
Action Statements and Completion Times to TS 3.7.12 and modification
of Condition D to address the filtration function of ECCS PREACS
does not impact the accident analysis or associated assumptions. The
new conditions only address actions to be taken when portions of
ECCS PREACS (an accident mitigation system) is out-of-service.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The proposed new conditions recognize that
there may be limited leakage situations when filtration is not
required to meet the accident analysis assumptions. Allowing safety
equipment to be inoperable while it is not required is not reducing
the analyzed margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond,
Virginia 23219.
NRC Branch Chief: Gloria J. Kulesa.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: October 16, 2009.
Description of amendment request: The license amendment request
(LAR) adds two references to the list of NRC approved methodologies
contained in the Technical Specifications (TSs). Specifically,
Westinghouse document WCAP-8745-P-A, ``Design Bases for Thermal
Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function,''
and the Dominion Fleet Report DOM-NAF-2-A, ``Reactor Core Thermal-
Hydraulics Using the VIPRE-D Computer Code,'' including Appendix B,
``Qualification of the Westinghouse WRB-1 CHF [Critical Heat Flux]
Correlation in the Dominion VIPRE-D Computer Code,'' in TS 6.2.C as a
referenced analytical methodology report.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Approval of the proposed changes will allow Dominion to use the
VIPRE-D/WRB-1 and VIPRE-D/W-3 code/correlation pairs to perform
licensing calculations of Westinghouse 15x15 Upgrade fuel in Surry
cores, using the DDLs [Deterministic Design Limits] documented in
Appendix B of the DOM-NAF-2-A Fleet Report and the SDL [Statistical
Design Limit]. Neither the code/correlation pair nor the Statistical
Departure from Nucleate Boiling Ratio (DNBR) Evaluation Methodology
make any contribution to the potential accident initiators and thus
cannot increase the probability of any accident. Further, since both
the deterministic and statistical DNBR limits meet the required
design basis of avoiding Departure from Nucleate Boiling (DNB) with
95% probability at a 95% confidence level, the use of the new code/
correlation and the Statistical DNBR Evaluation Methodology do not
increase the potential consequences of any accident. Finally, the
full core DNB design limit provides increased assurance that the
consequences of a postulated accident which includes radioactive
release would be minimized because the overall number of rods in DNB
would not exceed the 0.1% level. The pertinent evaluations to be
performed as part of the cycle specific reload safety analysis to
confirm that the existing safety analyses remain applicable have
been performed and determined to be acceptable. The use of a
different code/correlation pair will not increase the probability of
an accident because plant systems will not be operated in a
different manner, and system interfaces will not change. The use of
the VIPRE-D/WRB-1 and VIPRE-D/W-3 code/correlation pairs to perform
licensing calculations of Westinghouse 15x15 Upgrade fuel in Surry
cores will not result in a measurable impact on normal operating
plant releases and will not increase the predicted radiological
consequences of accidents postulated in the UFSAR [Updated Final
Safety Analysis Report].
The remaining proposed changes are being made to enhance the
completeness of the Surry TS and to achieve consistency with NUREG-
1431 Rev. 3. The proposed changes do not add or modify any plant
systems, structures or components (SSCs). The proposed changes to
relocate TS parameters to the COLR [Core Operating Limits Report]
are programmatic and administrative in nature. These changes do not
physically alter safety-related systems nor affect the way in which
safety-related systems perform their functions. Additional Safety
Limits on the DNB design basis and peak fuel centerline temperature
are being imposed in TS 2.1, ``Safety Limit, Reactor Core,'' and the
Reactor
[[Page 62839]]
Core Safety Limits figure is being relocated to the COLR. The
additional Safety Limits are consistent with the values stated in
the UFSAR and those being proposed herein. The proposed changes do
not, by themselves, alter any of the relocated parameter limits. The
removal of the cycle-specific parameter limits from the TS does not
eliminate existing requirements to comply with the parameter limits.
TS 6.2.C continues to ensure that the analytical methods used to
determine the core operating limits meet NRC reviewed and approved
methodologies and that applicable limits of the safety analyses are
met. Deletion of the obsolete limits associated with N-1 loop
operation (TS 2.1.A.2, TS 2.1.A.3, TS Figure 2.1-2, TS Figure 2.1-3)
and fuel densification (TS figure 2.1-4) is acceptable since these
limits no longer represent limiting conditions for operation and are
not required to be in the Technical Specifications.
Thus, the proposed changes do not affect initiators of analyzed
events or assumed mitigation of accident or transient events.
Therefore, this 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?
Response: No.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
The use of VIPRE-D and its applicable fuel design limits for
DNBR does not impact any of the applicable design criteria and all
pertinent licensing basis criteria will continue to be met.
Demonstrated adherence to these standards and criteria precludes new
challenges to components and systems that could introduce a new type
of accident. Setpoint safety analysis evaluations have demonstrated
that the use of VIPRE-D is acceptable. Design and performance
criteria will continue to be met and no new single failure
mechanisms will be created. The use of the VIPRE-D code/correlation
or the Statistical DNBR Evaluation Methodology does not involve any
alteration to plant equipment or procedures that would introduce any
new or unique operational modes or accident precursors.
The proposed change adds a new surveillance requirement of RCS
[Reactor Coolant System] Total Flow Rate and requests the addition
of an already approved method for determining plant operating
limits. The proposed change does not adversely affect accident
initiators or precursors, nor does it alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of SSCs to perform their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes to relocate TS parameters to the COLR are
programmatic and administrative in nature. Additional Safety Limits
on the DNB design basis and peak fuel centerline temperature are
being imposed in TS 2.1, ``Safety Limit, Reactor Core,'' and the
Reactor Core Safety Limits figure is being relocated to the COLR.
The additional Safety Limits are consistent with the values stated
in the UFSAR and those being proposed herein.
Approval of the proposed changes will allow Dominion to use the
VIPRE-D/WRB-1 and VIPRE-D/W-3 code/correlation pairs to perform
licensing calculations of Westinghouse 15x15 Upgrade fuel in Surry
cores, using the DDLs documented in Appendix B of the DOM-NAF-2-A
Fleet Report and the SDL documented herein. The SDL has been
developed in accordance with the Statistical DNBR Evaluation
Methodology. The DNBR limits meet the design basis of avoiding DNB
with 95% probability at a 95% confidence level. The use of the
VIPRE-D/WRB-1 code/correlation provides the same margin to safety as
the current code/correlation COBRA/WRB-1 used at Surry.
Therefore, the proposed TS 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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Gloria Kulesa.
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].
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: April 17, 2009.
Brief description of amendment: The amendment revises Operating
License No. DPR-49 by changing ``FPL Energy Duane Arnold, LLC'' to
``NextEra Energy Duane Arnold, LLC,'' where appropriate, to reflect the
renaming of FPL Energy Duane Arnold, LLC to NextEra Energy Duane
Arnold, LLC.
Date of issuance: November 13, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 275.
Facility Operating License No. DPR-49: The amendment revised the
License and Appendix B--Additional Conditions.
Date of initial notice in Federal Register: June 30, 2009 (74 FR
31324).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 13, 2009.
No significant hazards consideration comments received: No.
[[Page 62840]]
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 2, 2009.
Brief description of amendment: The amendment (1) deleted Technical
Specification (TS) surveillance requirement (SR) 3.1.3.2 and revised SR
3.1.3.3, (2) removed reference to SR 3.1.3.2 from Required Action A.3
of TS 3.1.3, ``Control Rod OPERABILITY,'' and (3) revised Example 1.4-3
in Section 1.4, ``Frequency,'' to clarify the applicability of the 1.25
surveillance test interval extension. The changes are in accordance
with NRC-approved TS Task Force (TSTF) traveler TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM [Source Range Monitor]
Insert Control Rod Action.''
Date of issuance: November 12, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 235.
Facility Operating License No. DPR-46: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 30, 2009 (74 FR
31325).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 12, 2009.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, (SSES Units 1 and 2) Luzerne County,
Pennsylvania
Date of application for amendments: March 24, 2009, as supplemented
by letters dated April 24, and September 11, 2009.
Brief description of amendments: The change revised the allowable
value in the Technical Specification (TS) Table 3.3.5.1-1 (Function
3.d) for the high-pressure coolant injection automatic pump suction
transfer from the condensate storage tank (CST) to the suppression
pool. The present allowable value for this transfer is greater than or
equal to 36 inches above the CST bottom. The change is to increase the
allowable value for this transfer to occur at greater than or equal to
40.5 inches above the CST bottom.
Additionally, the amendment also included an editorial/
administrative change which corrected a typographical error in the SSES
Units 1 and 2 TS Section 3.10.8.f.
Date of issuance: November 9, 2009.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 254 for Unit 1 and 234 for Unit 2.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: October 6, 2009, (74 FR
51332).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 9, 2009.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: July 28, 2009, supplemented by
letters dated September 16 and 30, 2009.
Brief Description of amendments: These amendments revise the
Technical Specifications (TS) of Surry Power Station, Units 1 and 2.
The request proposed changes to the inspection scope and repair
requirements of TS Section 6.4.Q, ``Steam Generator (SG) Program,'' to
the reporting requirements of TS Section 6.6.A.3, ``Steam Generator
(SG) Tube Inspection Report,'' and to TS Sections 4.13 and 3.1.C, ``RCS
[Reactor Coolant System] Operational Leakage.'' The proposed changes
would establish alternate repair inspection and criteria for portions
of the SG tubes within the tubesheet. The alternate inspection and
repair criteria would be applicable to Unit 1 during Refueling Outage
23 (fall 2010) and the subsequent operating cycle and to Unit 2 during
Refueling Outage 22 (fall 2009) and the subsequent operating cycle.
Date of issuance: November 5, 2009.
Effective date: Unit 1 is effective as of its date of issuance and
shall be implemented by the end of the fall 2010 refueling outage. Unit
2 is effective as of its date of issuance and shall be implemented by
the end of the fall 2009 refueling outage.
Amendment Nos.: 267 and 266.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the licenses and the technical specifications.
Date of initial notice in Federal Register: August 19, 2009 (74 FR
41939).
The supplements dated September 16, 2009 and September 30, 2009,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated November 5, 2009.
No significant hazards consideration comments received: No.
Dated at Rockville, MD, this 19th day of November 2009.
For The Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-28630 Filed 11-30-09; 8:45 am]
BILLING CODE 7590-01-P