[Federal Register Volume 72, Number 223 (Tuesday, November 20, 2007)]
[Notices]
[Pages 65360-65377]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-22331]
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NUCLEAR REGULATORY COMMISSION
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 staff) 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 October 25, 2007, to November 7, 2007. The
last biweekly notice was published on November 6, 2007 (72 FR 62685).
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 10 CFR 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. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
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.
[[Page 65361]]
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, 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 delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice,
person(s) may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
via electronic submission through the NRC E-Filing system for a hearing
and a petition for leave to intervene. 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 01F21, 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 within 60
days, 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 set forth 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 supports 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 participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, 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.
A request for hearing or a petition for leave to intervene 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 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 a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must 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 Viewer
TM to access the Electronic Information Exchange (EIE), a
component of the E-Filing system.
The Workplace Forms Viewer TM 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
[[Page 65362]]
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 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 technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, 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 petition
and/or request 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). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
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, the 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. 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.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this amendment action, 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 01F21, 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. 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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland.
Date of amendments request: October 17, 2007.
Description of amendments request: The proposed amendment would
modify the Technical Specifications (TS) to establish more effective
and appropriate action, surveillance, and administrative requirements
related to the inoperability of snubbers in accordance with Nuclear
Regulatory Commission (NRC)-approved TS Task Force (TSTF) change
traveler TSTF-372-A, Revision 4. Specifically, the proposed amendment
would add Limiting Condition for Operation (LCO) 3.0.8. The NRC staff
issued a ``Notice of Opportunity To Comment on Model Safety Evaluation
on Technical Specification Improvement To Modify Requirements Regarding
the Addition of LCO 3.0.8 on the Inoperability of Snubbers Using the
Consolidated Line Item Improvement Process'' in the Federal Register on
November 24, 2004 (69 FR 68412). The notice included a model safety
evaluation (SE) and a model no-significant-hazards-consideration (NSHC)
determination. The NRC staff issued a ``Notice of Availability of Model
Application Concerning Technical Specification Improvement To Modify
Requirements Regarding the Addition of Limiting Condition for Operation
3.0.8 on the Inoperability of Snubbers Using the Consolidated Line Item
Improvement Process'' in the Federal Register on May 4, 2005 (70 FR
23252). The notice included a model application, including a revised
model SE. In its application dated October 17, 2007, 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 adopted by the licensee is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO 3.0.8
are no different than the consequences of an accident while relying
on the TS required actions in effect without the allowance provided
by proposed LCO 3.0.8. Therefore, the consequences of an accident
previously evaluated are not significantly affected by this change.
The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
[[Page 65363]]
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's assessment and management of plant
risk. The net change to the margin of safety is insignificant.
Therefore, this change does not involve a significant reduction in a
margin of safety.
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 amendments request involves NSHC.
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: Mark G. Kowal.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: September 14, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) requirements related to
control room envelope habitability. The proposed changes include
revisions to the control room post-accident recirculation system, the
instrument operating conditions for isolation functions, and a control
room envelope habitability program. The changes are consistent with TS
Task Force (TSTF) Change Traveler TSTF-448-A, Revision 3, ``Control
Room Habitability,'' except for the differential pressure surveillance
requirements. The availability of this TS improvement was published in
the Federal Register on January 17, 2007 (72 FR 2022).
In addition to the changes related to TSTF-448-A, the proposed
amendment would: (1) Align TS with those delineated in NUREG-1431,
Revision 3, ``Standard Technical Specifications, Westinghouse Plants,''
to the extent necessary to adopt TSTF-448-A, including the adoption of
the necessary portions of TSTF-51-A, Revision 2, ``Revise Containment
Requirements During Handling of Irradiated Fuel and Core Alterations,''
and TSTF-287-A, Revision 5, ``Ventilation System Envelope Allowed
Outage Time,'' (2) add TS for control room radiation monitor R-23
(ventilation system air monitor), and (3) reformat or clarify current
TS.
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?
No.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility. The proposed changes do not prevent
the ability of structures, systems, and components (SSCs) to perform
their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits. This is a
revision to the TS for the control room post-accident recirculation
system and control room isolation function, which are mitigation
systems designed to minimize unfiltered air in-leakage into the
control room envelope and to filter the control room envelope
atmosphere to protect the control room envelope occupants following
accidents previously analyzed. An important part of the system is
the control room envelope boundary. The control room envelope post-
accident recirculation system is not an initiator or precursor to
any accident previously evaluated. Therefore, the probability of any
accident previously evaluated is not significantly increased.
Establishing operability requirements for SSCs, performing tests
and implementing programs that verify the integrity of the control
room envelope boundary and control room envelope habitability ensure
that the mitigation features are capable of performing their assumed
functions. Therefore, 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. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No.
The proposed changes will not significantly change the
requirements of the control room envelope ventilation system or its
function during accident conditions. No new or different accidents
result from performing the new surveillance or following the new
program. 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 significant change in the methods governing normal
plant operation. The proposed changes are consistent with the safety
analysis assumptions including the revised gas decay tank and volume
control tank rupture analysis and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No.
The proposed changes do 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 these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis for an unacceptable period without compensatory measures. The
proposed changes do not significantly affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, Riverside 2, Richmond,
VA 23219.
NRC Acting Branch Chief: Travis L. Tate.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: October 2, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Sections 3.7, ``Auxiliary
Electrical Systems'' and 4.6, ``Periodic Testing of Emergency Power
System,'' to change the testing requirements for ensuring operability
of the remaining operable emergency diesel generator (EDG) when the
other EDG is inoperable. In addition, the
[[Page 65364]]
proposed amendment would add a new specification when two EDGs are
inoperable and revise the surveillance requirements for the EDGs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No.
The proposed amendment would clarify testing requirements for
the operable EDG, when one EDG is inoperable, and limit testing to
only the intended purpose of the requirement. The intended purpose
of the testing requirement is to provide reasonable assurance that
when an EDG is inoperable, the opposite EDG is operable. The
proposed change does not affect the initiators of analyzed events or
the assumed mitigation of accident or transient events.
Specifically, testing of the remaining operable diesel will still
occur unless evaluation of the inoperable EDG confirms that its
failure is not attributable to a common cause failure mechanism.
Furthermore, the proposed change clarifies the surveillance testing
necessary to give reasonable assurance of operability and restricts
the amount of time to perform the testing (i.e. with two inoperable
EDGs) to two hours. This ensures no significant increase in the
probability of a loss-of-power during the period of the confirming
surveillance concurrent with an opposite train inoperable EDG.
Elimination of unnecessary testing by acceptable evaluation of the
operable EDG reduces component wear and promotes overall EDG
reliability and availability. Clarification of required testing and
restriction in the amount of time to complete the surveillance to
confirm operability, reduces the probability and significance of
common mode failures.
The proposed amendment would also add a new specification
allowing two EDGs to be inoperable for up to two hours. This change
does not significantly increase the initiators of analyzed events or
the assumed mitigation of any accidents or transients. Therefore,
the proposed amendment does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not involve a physical alteration of
the plant or a change in the methods used to respond to any
evaluated plant accident. No new or different equipment is being
installed and no installed equipment is being removed or operated in
a different manner. Only a surveillance test clarification and
limited two-hour action statement have been added to permit testing
of the opposite train, operable EDG. Although the diesel generators
will be tested in a different manner, the proposed changes will
improve the availability and reliability of the diesel generators
without creating the possibility of a new or different kind of
accident from any accident previously evaluated. Furthermore, there
is no alteration to the parameters within which the plant is
normally operated or in the setpoints, which initiate protective or
mitigative actions. Since the diesel generators will continue to be
operated in the same manner and the proposed test protocol will
improve diesel generator availability and reliability, no new
failure modes are introduced by the proposed 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 would add a TS allowing two EDGs to be
inoperable for up to two hours before the plant must be shut down in
a controlled manner. Allowing two EDGs to be inoperable for this
limited period of time, while the normal offsite power source
remains available, is consistent with Regulatory Guide 1.93 and not
considered to be a significant reduction in a margin of safety.
Station operations and EDG surveillance requirements are not
adversely affected by the proposed change. Furthermore, the proposed
amendment does not adversely impact the condition or performance of
structures, systems or components relied upon for accident
mitigation or any safety analysis assumptions. The proposed
amendment adds provisions to reduce EDG wear and increase
availability.
Therefore, the proposed amendment to the KPS [Kewaunee Power
Station] TS 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., Counsel for Dominion Energy Kewaunee, Inc.,
120 Tredegar Street, Richmond, VA 23219.
NRC Acting Branch Chief: Travis L. Tate.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: October 16, 2007.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to accommodate plant modifications
that will address water hammer concerns described in Generic Letter 96-
06, ``Assurance of Equipment Operability and Containment Integrity
During Design-Basis Conditions,'' dated September 30, 1996.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The requested license amendment seeks approval for the Low
Pressure Service Water Reactor Building Waterhammer Prevention
System that is being added to the design of the three Oconee Units
and the associated revised Technical Specifications. The Low
Pressure Service Water Reactor Building Waterhammer Prevention
modification will provide a combination passive and automatic means
to isolate the Low Pressure Service Water flow stream to the Reactor
Building Cooling Units, Reactor Building Auxiliary Coolers, and
Reactor Coolant Pump Motor Coolers on a loss of Low Pressure Service
Water flow that can lead to a waterhammer should the Low Pressure
Service Water system become depressurized.
New check valves and air operated valves are added into an
Engineered Safeguards flowpath. The existing Low Pressure Service
Water header that discharges from the Reactor Building Cooling Units
is to be modified by separating it into two headers and then joining
back into a common header. Each header will contain two air operated
valves. The Waterhammer Prevention System maintains the Low Pressure
Service Water System inside containment water solid during a Loss of
Offsite Power such that voids, which could later collapse, cannot
form. The Waterhammer Prevention System will eliminate an Operable
but degraded/non-conforming condition associated with waterhammers.
The design of the proposed modification and its associated
Technical Specifications will provide means to assure that the Low
Pressure Service Water Reactor Building Waterhammer Prevention
System operates at a performance level necessary to provide for safe
operation of the Low Pressure Service Water system following
installation on each of the three Units. The system is designed such
that a single active failure will not prevent the system from
preventing a waterhammer event if power is lost to the Low Pressure
Service Water pumps (e.g., Loss of Offsite Power), nor will a single
active failure prevent the Engineered Safeguards flowpath from being
available if needed during a Loss of Coolant Accident or Main Steam
Line Break. Evaluations have been performed to assure that the risk
of adding new hardware is acceptable.
Therefore, the addition of this modification and associated
Technical Specifications does not significantly increase the
probability or consequences of any accident previously evaluated.
[[Page 65365]]
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed Low Pressure Service Water Reactor Building
Waterhammer Prevention Modification and its associated Technical
Specifications will provide a means to assure the mechanical and
electrical components operate at a performance level necessary to
provide for safe operation of the modified Low Pressure Service
Water system flow to the Reactor Building Cooling Units, Reactor
Building Auxiliary Coolers and Reactor Coolant Pump Motor Coolers.
The change enhances the plant design by eliminating the
possibility of significant waterhammers that occur on a loss of Low
Pressure Service Water flow to the above components.
The modification does not add any new single active failures
that would prevent the Low Pressure Service Water System from
supplying cooling water to the Reactor Building Cooling Units. The
Reactor Building Cooling Units will be isolated briefly during an
Engineered Safeguards event; however, the flow path will be restored
before cooling is required following the event. Since cooling was
previously not available until after power restoration following a
Loss of Offsite Power, there is no change in system response
regarding Low Pressure Service Water flow through the Reactor
Building Cooling Units when compared to the previous design.
Therefore, the proposed modification and associated Technical
Specifications will not create the possibility of a new or different
kind of accident from any kind of accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change does not adversely affect any plant safety
limits, setpoints, or design parameters. The change also does not
adversely affect the fuel, fuel cladding, Reactor Coolant System, or
Containment Operability. The Reactor Building Cooling Units will be
isolated briefly during an Engineered Safeguards event; however, the
flow path will be restored before cooling is required following the
event.
Since cooling is currently not available until after power
restoration following a Loss of Offsite Power, there is no change in
system response regarding Low Pressure Service Water flow through
the Reactor Building Cooling Units when compared to the previous
design.
The modification mitigates significant waterhammers in the Low
Pressure Service Water piping to the Reactor Building Cooling Units
and Reactor Cooling Pump Motor Coolers. The change will maintain the
ability to provide Low Pressure Service Water flow to safety related
loads following Loss of Offsite Power events.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina.
Date of amendment request: October 22, 2007.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to accommodate the use of AREVA NP
Mark-B-HTP fuel.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revisions to the technical specifications and to
Duke's NRC-approved methodology reports support the use of the AREVA
NP Mark-B-HTP fuel design. The methodology will be approved by the
NRC prior to plant operation with the new fuel. The proposed safety
limit ensures that fuel integrity will be maintained during normal
operations and anticipated operational transients. The core
operating limits report will be developed in accordance with the
approved methodology. The proposed safety limit value does not
affect the performance of any equipment used to mitigate the
consequences of an analyzed accident. There is no impact on the
source term or pathways assumed in accidents previously assumed. No
analysis assumptions are violated and there are no adverse effects
on the factors that contribute to offsite or onsite dose as the
result of an accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed safety limit value does not change the methods
governing normal plant operation, nor are the methods utilized to
respond to plant transients altered. The BHTP correlation is not an
accident/event initiator. No new initiating events or transients
result from the use of the BHTP correlation or the related safety
limit change.
3. Involve a significant reduction in a margin of safety.
The proposed safety limit value has been established in
accordance with the methodology for the BHTP correlation to ensure
that the applicable margin of safety is maintained (i.e. there is at
least 95% probability at a 95% confidence level that the hot fuel
rod does not experience DNB). The other reactor core safety limits
will continue to be met by analyzing the reload using NRC approved
methods and incorporation of resultant operating limits into the
Core Operating Limits Report (COLR).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of amendment request: August 30, 2007.
Description of amendment request: The proposed amendment would
modify Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2)
Technical Specification (TS) requirements related to control room
envelope habitability in TS 3.7.10, ``Control Room Emergency
Ventilation System (CREVS)'' and TS Section 5.5, ``Administrative
Controls--Programs and Manuals.'' This change is consistent with
Nuclear Regulatory Commission (NRC)-approved Technical Specification
Task Force (TSTF) Change Traveler TSTF-448, Revision 3. The
availability of this TS revision was announced in the Federal Register
on January 17, 2007 (72 FR 2022) as part of the consolidated line item
improvement process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration adopted by the licensee is
presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change does not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility. The proposed change does not alter or
prevent the ability of structures, systems, and components (SSCs) to
perform their intended function to mitigate the consequences of an
initiating event within the assumed
[[Page 65366]]
acceptance limits. The proposed change revises the TS for the CRE
emergency ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency ventilation
system is the CRE boundary. The CRE emergency ventilation system is not
an initiator or precursor to any accident previously evaluated.
Therefore, the probability of any accident previously evaluated is not
increased. Performing tests to verify the operability of the CRE
boundary and implementing a program to assess and maintain CRE
habitability ensure that the CRE emergency ventilation system is
capable of adequately mitigating radiological consequences to CRE
occupants during accident conditions, and that the CRE emergency
ventilation system will perform as assumed in the consequence analyses
of design basis accidents. Thus, the consequences of any accident
previously evaluated are not increased. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 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 does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new surveillance
or following the new program. The proposed change does not involve a
physical alteration of the plant (i.e., no new or different type of
equipment will be installed) or a significant change in the methods
governing normal plant operation. The proposed change does not alter
any safety analysis assumptions and is consistent with current plant
operating practice. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. 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.
Therefore, the proposed change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the 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: David W. Jenkins, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: Mark G. Kowal.
FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: October 1, 2007.
Description of amendment request: The proposed amendments would
revise the accident source term in the design-basis radiological
consequences analyses and the associated Technical Specifications
(TSs), pursuant to Section 50.67 of Part 50 of Title 10 of the Code of
Federal Regulations (10 CFR 50.67). The proposed amendments would
revise the licensing basis of Point Beach Nuclear Plant, Units 1 and 2
(PBNP) to support a full-scope application of an Alternative Source
Term (AST) methodology. The AST methodology will modify PBNP's
licensing bases by: (1) Replacing the current accident source term with
an AST as described in 10 CFR 50.67 for design-basis accidents (DBA)
radiological consequences, and (2) establishing the 10 CFR 50.67 Total
Effective Dose Equivalent (TEDE) dose limits as acceptance criteria for
the radiological consequences of DBAs.
TS changes associated with the AST methodology change are: TS 1.1,
a reduction in the definition of the maximum allowable containment leak
rate. TS 3.4.16, the specific activity of the reactor coolant is
revised for dose equivalent iodine. TS 3.7.9, a new mode of operation
for the Control Room Emergency Filtration System (CREFS), which will
allow operation of the CREFS with filtered outside and filtered
recirculated air.
TS 3.7.13, the specific activity of the secondary coolant is
revised for dose equivalent iodine. In addition, a modification to the
residual heat removal system, containment spray and their support
systems, will be made to support operation of the containment spray
system during containment spray recirculation.
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 results of the applicable radiological design basis accident
(DBA) re-evaluation demonstrated that, with the requested changes,
the dose consequences of these limiting events are within the
regulatory limits and guidance provided by the NRC in 10 CFR 50.67
and Regulatory Guide 1.183 for alternative source term (AST)
methodology. The AST is an input to calculations used to evaluate
the consequences of an accident and does not by itself affect the
plant response or the actual pathway of the activity released from
the fuel. It does, however, better represent the physical
characteristics of the release such that appropriate mitigation
techniques may be applied.
The change from the original source term to the new proposed AST
is a change in the analysis method and assumptions and has no effect
on accident initiators or causal factors that contribute to the
probability of occurrence of previously analyzed accidents. Use of
an AST to analyze the dose effect of DBAs shows that regulatory
acceptance criteria for the new methodology continues to be met.
Changing the analysis methodology does not change the sequence or
progression of the accident scenario.
The proposed Technical Specification changes reflect the plant
configuration that will support implementation of the AST analyses.
The equipment affected by the proposed changes is mitigative in
nature and relied upon after an accident has been initiated. The
operation of various filtration systems, the residual heat removal
and the containment spray system, including associated support
systems, has been considered in the evaluations for these proposed
changes. While the operation of these systems does change with the
implementation of an AST, the affected systems are not accident
initiators, and application of the AST methodology itself, is not an
initiator of a design basis accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 65367]]
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
As described in Item 1 above, the changes proposed in this
license amendment request involve the use of a new analysis
methodology and related regulatory acceptance criteria. The proposed
Technical Specification changes reflect the plant configuration that
will support implementation of the new methodology. No new or
different accidents result from utilizing the proposed changes.
Although the proposed changes require modifications to the control
room emergency ventilation system, as well as modifications to the
residual heat removal system and containment spray system, these
changes will not initiate a new or different kind of accident since
they are related to system capabilities that provide protection from
accidents that have already occurred. As a result, no new failure
modes are being introduced that could lead to different accidents.
These changes do not alter the nature of events postulated in the
Updated Final Safety Analysis Report nor do they introduce any
unique precursor mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of Safety.
Response: No.
As described in Item 1, the changes proposed in this license
amendment involve the use of a new analysis methodology and related
regulatory acceptance criteria. The proposed Technical Specification
changes reflect the plant configuration that will support
implementation of the new methodology. Safety margins and analytical
conservatisms have been evaluated and have been found to be
acceptable. The analyzed events have been carefully selected and,
with plant modifications, margin has been retained to ensure that
the analyses adequately bound postulated event scenarios. The
proposed changes continue to ensure that the dose consequences of
DBAs at the exclusion area and low population zone boundaries and in
the control room are within the corresponding acceptance criteria
presented in RG 1.183 and 10 CFR 50.67. The margin of safety for the
radiological consequences of these accidents is provided by meeting
the applicable regulatory limits, which are set at or below the 10
CFR 50.67 limits. An acceptable margin of safety is inherent in
these limits.
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. Antonio Fernandez, Senior Attorney, FPL
Energy Point Beach, LLC P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Acting Branch Chief: Travis L. Tate.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: September 27, 2007.
Description of amendment request: The proposed amendment would
revise the operability requirements contained in Technical
Specification (TS) Section 3.2.7, ``Reactor Coolant System Isolation
Valves,'' and associated requirements contained in TS Section 3.6.2,
``Protective Instrumentation.'' The proposed changes would modify the
conditions for which reactor coolant system isolation valves (RCSIVs)
and associated isolation instrumentation must be operable to include
the hot shutdown reactor operating condition (i.e., when fuel is in the
reactor vessel and the reactor coolant temperature is greater than 212
[deg]F). In addition, new requirements are proposed to require that the
RCSIVs in the shutdown cooling (SDC) system and associated isolation
instrumentation be operable during the cold shutdown reactor operating
condition (fuel is in the reactor vessel and the reactor coolant
temperature is less than or equal to 212 [deg]F) and the refueling
reactor operating condition (i.e., when fuel is in the reactor vessel
and the reactor coolant temperature is less than 212 [deg]F). These
proposed changes will require operability of RCSIVs during conditions
other than the power operating condition, and are similar in concept to
primary containment isolation valve operability requirements contained
in NUREG-1433, ``Standard Technical Specifications General Electric
Plants, BWR/4.'' Lastly, TS Section 3.6.2 (Table 3.6.2b) would be
revised to delete unnecessary operability requirements for the cleanup
system and SDC system high area temperature isolation instrumentation,
consistent with the proposed revisions to the RCSIV operability
requirements.
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 provide more stringent requirements for
operation of NMP1. These include requiring operability of RCSIVs and
associated isolation instrumentation during the hot shutdown
condition and requiring RCSIVs in the SDC system and associated
instrumentation to be operable during the cold shutdown and
refueling operating conditions. Requiring RCSIV operability during
the hot shutdown operating condition ensures that reactor coolant
loss in the event of a rupture of a line connected to the reactor
coolant system (RCS) is minimized, and the release of radioactive
material to the environment is consistent with the assumptions used
in the analyses for design basis accidents. Requiring operability of
the RCSIVs in the SDC system during the cold shutdown and refueling
operating conditions provides protection against potential draining
of the reactor vessel through the SDC system during shutdown
conditions, which is when the SDC system is normally operated.
In addition, operability requirements for the cleanup system and
SDC system high area temperature isolation instrumentation are
revised to be consistent with the proposed revisions to the RCSIV
operability requirements and with NUREG-1433. The high area
temperature isolation instrumentation need not be operable in the
cold shutdown and refueling conditions, since the probability and
consequences of design basis accidents are reduced due to the
pressure and temperature limitations of these operating conditions.
Also, system isolation on high area temperature would likely not
occur in the event of system leakage or line break since RCS
temperature during the cold shutdown and refueling conditions is
typically maintained below the high area temperature isolation
setpoints (190[deg]F for the cleanup system area and 170[deg]F for
the SDC system area).
The revised operability requirements for the RCSIVs and
associated isolation instrumentation do not result in operation that
would make an accident more likely to occur and do not alter
assumptions relative to mitigation of a previously evaluated
accident. Therefore, the 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 changes to the TS operability requirements for the
RCSIVs and associated isolation instrumentation do not alter or
involve any design basis accident initiators. The proposed changes
do not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or changes in the
methods governing normal plant operation. The proposed changes do
impose different RCSIV operability requirements that are more
stringent than existing requirements, and incorporate RCSIV
isolation instrumentation operability requirements that are
consistent with the RCSIV requirements and with NUREG-1433. These
changes continue to be consistent with
[[Page 65368]]
the assumptions in the safety analyses and licensing basis.
Therefore, the proposed changes will 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 changes to the TS operability requirements for the
RCSIVs and associated isolation instrumentation ensure that RCSIV
closure will occur when required to mitigate the consequences of
design basis accidents. The proposed changes also ensure that SDC
system isolation can be accomplished to protect against potential
draining of the reactor vessel through the SDC system during
shutdown conditions, which is when the SDC system is normally
operated. The imposition of these revised RCSIV operability
requirements either has no impact on or increases the margin of
plant safety. The plant responses to accidents will not be adversely
affected, and the accident mitigation equipment will continue to
function as assumed in the accident analyses. Therefore, the
proposed changes do 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No, 2 (NMP2), Oswego County, New York
Date of amendment request: September 19, 2007.
Description of amendment request: The proposed amendment would
revise NMP2 Limiting Condition for Operation (LCO) 3.10.1 to expand its
scope to include provisions for temperature excursions greater than 200
[deg]F as a consequence of inservice leak and hydrostatic testing, and
as a consequence of scram time testing initiated in conjunction with an
inservice leak or hydrostatic test, while considering operational
conditions to be in Mode 4. This change is consistent with Nuclear
Regulatory Commission (NRC)-approved Revision 0 to Technical
Specification (TS) Task Force (TSTF) Change Traveler, TSTF-484, ``Use
of TS 3.10.1 for Scram Time Testing Activities.'' The availability of
this TS revision was announced in the Federal Register on October 27,
2006 (71 FR 63050) as part of the consolidated line item improvement
process. The licensee affirmed the applicability of the model no
significant hazards consideration determination in its application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration adopted by the licensee is
presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
Technical Specifications currently allow for operation at greater
than [200][deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact the
probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
Criterion 2: The proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
Technical Specifications currently allow for operation at greater
than [200][deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. No new
operational conditions beyond those currently allowed by LCO 3.10.1 are
introduced. 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
or eliminate any existing 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 change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
Technical Specifications currently allow for operation at greater
than [200][deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact any
margin of safety. Allowing completion of inspections and testing and
supporting completion of scram time testing initiated in conjunction
with an inservice leak or hydrostatic test prior to power operation
results in enhanced safe operations by eliminating unnecessary
maneuvers to control reactor temperature and pressure. Therefore, the
proposed change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: September 25, 2007.
Description of amendment request: The proposed amendment would
revise the MNGP licensing basis to incorporate the results of a revised
small-break loss-of-coolant accident (LOCA) analysis to determining the
Low Pressure Coolant Injection (LPCI) loop select logic minimum
detectable break area. This analysis showed that a small break, rather
than the current large recirculation line break LOCA, would become the
limiting accident with respect to peak cladding temperature (PCT). In
conjunction with this proposed new licensing basis analysis, the
licensee proposed to revise the Table 3.3.5.1-1 (regarding emergency
core cooling system instrumentation) of the Technical Specifications
(TS) as follows: (1) change the allowable value from the current 24
inch water column to 100 inch water column for Function 2.j,
``Recirculation Riser Differential Pressure--High (Break Detection);''
and (2) change the associated channel calibration frequency
Surveillance Requirement (SR) from a nominal 12-month to a 24-month
interval.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Part 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration (NSHC).
The NRC staff reviewed the licensee's analysis, and has performed its
own as follows:
1. Do the proposed changes involve a significant increase in the
probability or
[[Page 65369]]
consequences of an accident previously evaluated?
No. The proposed changes to the PCT licensing basis and the TS do
not involve a physical alteration of the plant, i.e., no design change
to plant system, and no new or different type of equipment will be
installed. The proposed PCT change is an analysis result which is
within regulatory acceptance limits, and the proposed TS changes
reflect the revised analysis. Thus, the proposed changes affect only
parameters assumed for certain analyses, but do not adversely affect
accident initiators, precursors, plant design, configuration, or the
manner in which the plant is operated and maintained. The proposed
changes do not adversely affect the ability of structures, systems and
components to perform their intended safety function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not affect the source term, containment
isolation capability, or radiological consequences of any accident
previously evaluated. Furthermore, the proposed changes do not increase
the types and the amounts of radioactive effluent that may be released,
and do not significantly increase individual or cumulative
occupational/public radiation exposures. Therefore, the proposed
changes do not involve a significant increase in the probability or
consequences of any accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed changes do not involve a physical altering of the
plant (i.e., no new or different type of equipment will be installed)
or a change in methods governing normal plant operation. The
requirements in the TS will continue to assure operation of the plant
within its design specifications and safety limits. Therefore, the
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in the
margin of safety?
No. The proposed amendment would only change the analysis of record
LOCA PCT, the allowed value of an instrument function, and its
associated SR frequency. There will be no modification of any TS
limiting condition for operation, no change to any limit on previously
analyzed accidents, no change to how previously analyzed accidents or
transients would be mitigated, no change in any methodology used to
evaluate consequences of accidents, and no change in any operating
procedure or process. Therefore, the proposed amendment does not entail
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
its own analysis and has found that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the proposed amendment involves no significant hazards
consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Clifford G. Munson.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 5, 2007.
Description of amendment request: The proposed amendment requests a
change to Technical Specification 3.7(1)ci, ``Emergency Power Periodic
Test,'' related to the surveillance testing of the Fort Calhoun Station
emergency diesel generators (DGs) to support a modification to the DG
start circuitry.
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 removal of the anticipatory (idle speed) diesel generator
(DG) start signal on a reactor protective system (RPS) reactor trip
does not adversely affect the design function of the DGs and thus is
not an initiator of any previously evaluated accidents.
No Updated Safety Analysis Report (USAR) accident analyses take
credit for the anticipatory (idle speed) DG start following a design
basis accident (DBA). The DGs provide emergency power to their
respective 4.16 KV [Kilovolt] buses and will continue to do so after
the proposed modification is installed. Upon the occurrence of an
undervoltage condition on the bus or an engineered safety features
(ESF) signal, the modification provides a full speed DG start to
achieve rated voltage and frequency. The safety function of the DGs
is not altered by the installation of the modification. The
associated Technical Specification (TS) change allows surveillance
testing to reflect the way that the DGs start and load onto their
respective buses following the modification.
Deletion of a footnote containing historical information
pertaining to a one-time surveillance interval extension and the
punctuation correction are administrative changes. These
administrative changes do not increase the probability or
consequences of any accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The removal of the anticipatory (idle speed) diesel generator
(DG) start signal on an RPS reactor trip does not adversely affect
the design function of the DGs and thus does not create the
possibility of a new or different kind of accident. There are no
USAR accident analyses which take credit for the anticipatory (idle
speed) DG start following a DBA. The DGs provide emergency power to
their respective 4.16 KV buses and will continue to do so after the
proposed modification is installed. Upon the occurrence of an
undervoltage condition on the bus or an ESF signal, the modification
provides a full speed DG start to achieve rated voltage and
frequency. The safety function of the DGs is not altered by the
installation of this modification. The associated TS change allows
surveillance testing to reflect the way that the DGs start and load
onto their respective buses following the modification.
Deletion of a footnote containing historical information
pertaining to a one-time surveillance interval extension and the
punctuation correction are administrative changes that do not create
the possibility of a new or different kind of accident from any
previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The removal of the anticipatory (idle speed) diesel generator
(DG) start signal on an RPS reactor trip does not adversely affect
the design function of the DGs and thus does not involve a
significant reduction in a margin of safety. There are no USAR
accident analyses which take credit for the anticipatory (idle
speed) DG start following a DBA. The DGs provide emergency power to
their respective 4.16 KV buses and will continue to do so after
installation of the proposed modification. Upon the occurrence of an
undervoltage condition on the bus or an ESF signal, the modification
provides a full speed DG start to achieve rated voltage and
frequency. The safety function of the DGs is not altered by the
installation of this modification. The associated TS change allows
surveillance testing to reflect the way that the DGs will start and
load onto their respective buses following the modification.
Deletion of a footnote containing historical information
pertaining to a one-time surveillance interval extension and the
[[Page 65370]]
punctuation correction are administrative changes that do not reduce
a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: October 12, 2007.
Description of amendment request: The proposed amendment would
modify the Fort Calhoun Station, Unit 1 design and licensing basis to
increase the shutdown cooling (SDC) system entry temperature from 300
[deg]F to 350 [deg]F (cold leg), and the SDC entry pressure from 250
psia to 300 psia (indicated at the pressurizer). Additionally, the
licensee proposes to change to the Updated Safety Analysis Report
(USAR) described design methodology applied to the SDC heat exchangers.
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 shutdown cooling (SDC) system provides flow to the reactor
during long term cooling mode following a large break loss-of-
coolant accident (LOCA). In addition, the SDC system can supply
cooled sump water to the high pressure safety injection (HPSI) pumps
for long term core cooling. The SDC system is also designed to
reduce the temperature of the reactor coolant system (RCS) from 300
[deg]F to refueling temperature within 24 hours and to maintain the
proper RCS temperature during refueling. As such, the SDC system is
not an initiator for any accident previously evaluated.
The proposed change to increase the SDC entry temperature from
300 [deg]F to 350 [deg]F affects the inputs to the analysis of the
Boron Dilution Incident.
However, re-analysis of this accident with the increased
temperature does not result in an increase in the probability of the
accident. The proposed increase in SDC system design and operating
temperature and pressure has been evaluated for affects on system
piping and components using appropriate codes and standards. The
proposed changes do not introduce any failure mechanisms that would
initiate a previously analyzed accident. Therefore, the proposed
change to uprate the SDC system entry conditions does not result in
a significant increase in the probability of a previously evaluated
accident.
The potential effect of the proposed change on the consequences
of a previously evaluated accident has been considered. Re-analysis
of the Boron Dilution Incident with the proposed increased SDC entry
temperature does not result in an increase in the consequences of
the accident.
In addition, although an increase in the SDC system leakage test
pressure is proposed, the leakage test acceptance criteria (i.e.,
maximum permitted leakage per hour) will not be affected. Therefore,
the limit on post-accident leakage to atmosphere from the SDC system
is unchanged. The proposed increase in SDC system design and
operating temperature and pressure does not affect the redundancy or
availability of the SDC system. The design functions of the system
are not affected by the proposed change. Therefore, the SDC system
will still be capable of performing the safety functions needed to
mitigate 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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change alters the SDC system entry conditions and
increases the system leakage test pressure. In the current design,
the SDC system has been excluded from consideration as a pipe
rupture initiator since it is not normally in operation. It is used
for plant shutdown and startup, and for accident mitigation. With
the proposed change, the operating modes of the system will not be
affected. The proposed change increases the RCS temperature and
pressure at which the SDC system can be placed in service during
shutdown (or removed from service during startup), but the RCS, SDC,
and other plant systems are not operated in a different manner. The
increased heat load on the component cooling water (CCW) system
resulting from normal operation of the SDC at increased SDC
temperatures has been evaluated. The increased normal operating heat
load has been determined to be bounded by the post-accident CCW heat
load. Any adjustments to the cooldown rate needed to accommodate the
increased SDC entry temperature will be performed using approved
procedures consistent with current practice and would not require
operating the plant in a different manner.
The RCS cooldown rate limitations in the Technical
Specifications (TS) are not affected by the proposed change. In
addition, adjustments of CCW heat loads to maintain required CCW
inlet temperatures for the SDC (Low Pressure Safety Injection
(LPSI)) pump coolers, when operating at the increased SDC entry
temperature, will be in accordance with plant procedures and within
existing system capabilities. The low temperature overpressurization
(LTOP) analysis has been revised for the proposed change. However,
there are no effects on existing LTOP setpoints or operating
limitations, other than the proposed change to TS 2.1.1(11)(b),
which states that the unit cannot be placed on shutdown cooling
until the RCS has been cooled to <= 350 [deg]F. The proposed change
in SDC operating limitations does not introduce the possibility of
new or different equipment malfunctions or accident precursors.
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 margins of safety are established through design parameters,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change increases the SDC system entry
conditions for plant shutdown, startup and following postulated
accidents, and the SDC system leakage test pressure. However, the
accident mitigation function and post-accident operation of the
system is not affected. The operating limits on temperature and
pressure will remain below the design temperature and pressure for
the system. The time interval for operator action after a postulated
boron dilution event with the SDC system in operation is reduced,
however, the available time remains greater than the minimum
required time interval of 15 minutes. The proposed change does not
affect any design or operating parameter or setpoint used in the
accident analyses to establish 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: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: October 15, 2007.
Description of amendment requests: The proposed amendments would
relocate all periodic surveillance frequencies from the technical
specifications (TS) and place the frequencies under licensee control in
[[Page 65371]]
accordance with a new program, the Surveillance Frequency Control
Program.
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 involves the relocation of various
surveillance test intervals from TSs to a licensee-controlled
program and is administrative in nature. The proposed change does
not involve the modification of any plant equipment or affect basic
plant operation. The proposed change will have no impact on any
safety related structures, systems or components. Surveillance test
intervals are not assumed to be an initiator of any analyzed event,
nor are they assumed in the mitigation of consequences of accidents.
The [Surveillance Requirements] themselves will be maintained in the
TS along with the applicable Limiting Conditions for Operation
(LCOs) and Action statements. The surveillances performed at the
intervals specified in the licensee-controlled program will assure
that the affected system or component function is maintained, that
the facility operation is within the Safety Limits, and that the
LCOs are met.
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 does not involve any physical alteration of
plant equipment and does not change the method by which any safety-
related structure, system, or component performs its function or is
tested. As such, no new or different types of equipment will be
installed, and the basic operation of installed equipment is
unchanged. The methods governing plant operation and testing remain
consistent with current safety analysis assumptions.
Therefore, the proposed change will 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 is administrative in nature, does not negate
any existing requirement, and does not adversely affect existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there are no changes being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by relocation of
the surveillance test intervals to a licensee-controlled program.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
?>The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Thomas G. Hiltz.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendments request: October 17, 2007.
Description of amendments request: The proposed amendment would
modify the Technical Specifications (TS) to establish more effective
and appropriate action, surveillance, and administrative requirements
related to the inoperability of snubbers in accordance with Nuclear
Regulatory Commission (NRC)-approved TS Task Force (TSTF) change
traveler TSTF-372-A, Revision 4. Specifically, the proposed amendment
would add Limiting Condition for Operation (LCO) 3.0.8. The NRC staff
issued a ``Notice of Opportunity To Comment on Model Safety Evaluation
on Technical Specification Improvement To Modify Requirements Regarding
the Addition of LCO 3.0.8 on the Inoperability of Snubbers Using the
Consolidated Line Item Improvement Process'' in the Federal Register on
November 24, 2004 (69 FR 68412). The notice included a model safety
evaluation (SE) and a model no-significant-hazards-consideration (NSHC)
determination. The NRC staff issued a ``Notice of Availability of Model
Application Concerning Technical Specification Improvement To Modify
Requirements Regarding the Addition of Limiting Condition for Operation
3.0.8 on the Inoperability of Snubbers Using the Consolidated Line Item
Improvement Process'' in the Federal Register on May 4, 2005 (70 FR
23252). The notice included a model application, including a revised
model SE. In its application dated October 17, 2007, 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 adopted by the licensee is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO 3.0.8
are no different than the consequences of an accident while relying
on the TS required actions in effect without the allowance provided
by proposed LCO 3.0.8. Therefore, the consequences of an accident
previously evaluated are not significantly affected by this change.
The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's assessment
[[Page 65372]]
and management of plant risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
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 amendments request involves NSHC.
Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Mark G. Kowal.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: October 18, 2007.
Description of amendment request: The proposed amendments to
Technical Specification Administrative Controls Section 5.3.1 would
revise the training and qualifying education and experience eligibility
requirements for certain unit staff positions to correspond to a
defined training program. The training program is based on National
Academy for Nuclear Training guidance documents (ACADs) as described in
the licensee's October 18, 2007, application. The proposed changes will
also replace a specific position title with a generic position title
for the senior individual in charge of Health Physics. An application
that addressed similar issues was previously submitted on October 30,
2006, and notice of that application was provided in the Federal
Register on July 17, 2007 (72 FR 39084). Due to certain changes in the
specifics of the October 18, 2007, application, from those proposed in
the October 30, 2006, application, the application is being renoticed
in its entirety. This notice supersedes the notice published in the
Federal Register on July 17, 2007.
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 to Technical Specifications Administrative
Controls Section 5.3.1 involves the use of a more generic
designation for the unit staff position responsible for Health
Physics without reducing the level of authority required for that
position. The proposed change also allows the flexibility to use an
accredited program for qualifying personnel to fill certain unit
staff positions as stipulated in Enclosure 1 [of October 18, 2007,
application], which represents an acceptable alternative to the
qualification requirements for these positions as currently
specified in the Technical Specifications. Since the proposed
changes are administrative in nature, they do not involve any
physical changes to any structures, systems, or components, nor will
their performance requirements be altered. The proposed changes also
do not affect the operation, maintenance, or testing of the plant.
Therefore, the response of the plant to previously analyzed
accidents will not be affected. Consequently, the proposed changes
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 previously evaluated?
Response: No.
The proposed changes to the Technical Specifications will have
no adverse impact on the overall qualification of the unit staff.
The use of a more generic designation for the unit staff position
responsible for Health Physics and the proposed addition [of] a
statement to Section 5.3.1 that will reference this letter and the
accreditation information for the positions stipulated in Enclosure
1 will allow the use of an accredited program that has been endorsed
by the NRC and will ensure the educational requirements and power
plant experience for each unit staff position are properly satisfied
and will continue to fulfill applicable regulatory requirements.
Also, since no change is being made to the design, operation,
maintenance, or testing of the plant, no new methods of operation or
failure modes are introduced by the proposed changes. Therefore, the
possibility of a new or different kind of accident from any
previously evaluated is not created.
3. Does the proposed change involve a significant decrease in
the margin of safety?
Response: No
The proposed changes to the Technical Specifications will have
no adverse impact on the onsite organizational features necessary to
assure safe operation of the plant. Lines of authority for plant
operation are unaffected by the proposed changes. Also, the adoption
of the more generic designation of the individual responsible for
Health Physics will reduce the regulatory burden of having to devote
limited resources to process a license amendment whenever a title
change for this position is implemented. Accordingly, this reduction
in regulatory burden and the proposed addition of a statement to
Section 5.3.1 that will reference this letter and the use of
accreditation information provided in Enclosure 1, will allow the
use of an accredited program endorsed by NRC to qualify certain unit
staff positions and will improve organizational flexibility without
compromising plant safety. Therefore, the proposed changes do not
involve a significant decrease 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Branch Chief: Evangelos C. Marinos.
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: August 28, 2007, as supplemented on
October 9, 2007.
Description of amendment request: The proposed amendments would
revise the ``Maximum Power Level'' in paragraph 2.C(1) of the Vogtle
Electric Generating Plant Facility Operating Licenses NPF-68 and NPF-81
for Unit 1 and Unit 2, respectively. In addition, the amendments would
revise the definition of ``Rated Thermal Power (RTP)'' in Technical
Specification 1.1 for both units to reflect the change to the Maximum
Power Level. The proposed change increases the RTP from 3565 MWt to
3625.6 MWt, resulting in an increase of 1.7% from the current reactor
output. This increase in reactor core power level is referred to as a
Measurement Uncertainty Recapture (MUR) power uprate.
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?
Operating License--Maximum Power Level and Technical Specification
1.1--Definition of Rated Thermal Power
The increase in Maximum Power Level and Rated Thermal Power
(RTP) does not involve a significant increase in the probability or
consequences of an accident previously evaluated, because operation
at the higher power level will not cause any design or analysis
acceptance criteria to be exceeded. As a result, structural and
functional
[[Page 65373]]
integrity of the plant systems is maintained. Power level is an
input assumption to the equipment design and accident analyses, but
it is not itself an initiator for any transient. Therefore, the
probability of occurrence of an accident previously evaluated is not
affected.
The radiological consequences of operation at the Measurement
Uncertainty Recapture (MUR) power uprate conditions have been
assessed. It was concluded that offsite dose predictions remain
within the acceptance criteria for each of the accidents affected.
Therefore, the consequences of an accident previously evaluated are
not increased.
Technical Specification 1.1--Definition of Dose Equivalent Iodine
The proposed change to the definition of dose equivalent iodine
(DEI) impacts the reactor coolant activity surveillance and
calculations of accident consequences and makes these activities
consistent with each other. Neither of these functions affects the
probability of any accident previously evaluated.
In order to support the MUR power uprate, the accidents
previously evaluated in the Updated Final Safety Analysis Report
(UFSAR) were re-analyzed. As part of this reanalysis, the dose
conversion factors (DCFs) were reviewed, and a consistent set of
DCFs was used for all re-analyses based on Federal Guidance Report
No. 11, as suggested by RIS 2001-19. The results of these re-
analyses continue to meet the acceptance limits as currently
described in the UFSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Technical Specification 3.3.1, Table 3.3.1-1, Function 16--P-9 Setpoint
The revised Power Range Neutron Flux P-9 permissive nominal
setpoint and allowable value do not involve a significant increase
in the probability or consequences of an accident previously
evaluated, because operation with these revised values will not
cause any design or analysis acceptance criteria to be exceeded. The
structural and functional integrity of any plant system is
unaffected. The P-9 permissive function is part of the transient
mitigation response and is not itself an initiator for any
transient. Therefore, the probability of occurrence of an accident
previously evaluated is not affected.
The changes to the P-9 nominal setpoint and allowable value do
not affect the integrity of the fission product barriers utilized
for the mitigation of radiological dose consequences as a result of
an accident. The change continues to ensure that the pressurizer
power operated relief valves (PORVs) are not challenged following a
turbine trip without a reactor trip which, in turn, minimizes the
potential for a release. There are no offsite dose predictions for
this transient. Since it has been determined that the transient
results are unaffected by the change to the P-9 nominal setpoint and
allowable value, it is concluded that the consequences of an
accident previously evaluated are not increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Operating License--Maximum Power Level and Technical Specification
1.1--Definition of Rated Thermal Power
The increase in Maximum Power Level and RTP does not create the
possibility of a new or different kind of accident from any
previously evaluated, because no new operating configuration is
being imposed that will create a new failure scenario, and no new
failure modes are being created for any plant equipment. System and
component design bases have been reviewed. The proposed change does
not have an adverse effect on safety-related systems or components
and does not challenge the integrity of any safety-related system.
Therefore, the types of accidents defined in the UFSAR continue to
represent the credible spectrum of events to determine safe plant
operation.
Technical Specification 1.1--Definition of Dose Equivalent Iodine
The proposed change to the definition of Dose Equivalent Iodine
(DEI) ensures the reactor coolant activity surveillances are
consistent with the assumptions for initial conditions used in the
accident analyses. The proposed change does not involve the addition
or modification of any plant equipment. Neither does it alter the
design, configuration or method of operation of the plant.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Technical Specification 3.3.1, Table 3.3.1-1, Function 16--P-9 Setpoint
The revised Power Range Neutron Flux P-9 permissive nominal
setpoint and allowable value do not create the possibility of a new
or different kind of accident from any previously evaluated, because
these changes do not affect accident initiation sequences. No new
operating configuration is being imposed by the P-9 nominal setpoint
and allowable value changes that will create a new failure scenario.
In addition, no new failure modes are being created for any plant
equipment. Therefore, the types of accidents defined in the UFSAR
continue to represent the credible spectrum of events to determine
safe plant operation.
3. Does the proposed change involve a significant decrease in a
margin of safety?
Operating License--Maximum Power Level and Technical Specification
1.1--Definition of Rated Thermal Power
The increase in Maximum Power Level and RTP does not involve a
significant reduction in a margin of safety, because power level is
one of the inherent assumptions that determine the safe operating
range defined by the accident analyses, which are in turn protected
by the Technical Specifications. The acceptance criteria for the
accident analyses are conservative with respect to the operating
conditions defined by the Technical Specifications. The engineering
reviews performed for the MUR power uprate confirmed that the
accident analyses criteria are met at the revised value of MPL and
RTP. Therefore, the adequacy of the revised Facility Operating
Licenses and Technical Specifications to maintain the plant in a
safe operating range is also confirmed, and the increase in MPL and
RTP do not involve a significant decrease in a margin of safety.
Technical Specification 1.1--Definition of Dose Equivalent Iodine
The proposed change to the definition of dose equivalent iodine
(DEI) has the potential to affect the dose consequences offsite and
in the control room. However, the results of the re-analyses of the
accidents previously evaluated demonstrate the dose consequences at
all locations remain within the regulatory acceptance limits, and
the margin of safety as defined by 10 CFR 100 and GDC 19 has not
been significantly reduced.
Technical Specification 3.3.1, Table 3.3.1-1, Function 16--P-9 Setpoint
The change to the P-9 nominal setpoint and allowable value does
not involve a significant reduction in a margin of safety because
the margin of safety associated with the P-9 setpoint, as verified
by the results of the applicable transient analyses, is within
acceptable limits. The adequacy of the revised Technical
Specification values to maintain the plant in a safe operating range
has been confirmed. Therefore, the change to the P-9 nominal
setpoint and allowable value does not involve a significant decrease
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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 27, 2007.
Description of amendment request: The amendments would revise the
licensee's fire protection program requirements as documented in the
licensee's Fire Hazard's Analysis Report. Specifically, the licensee
requests the use of reactor operator manual actions in lieu of meeting
protection requirements of circuit separation.
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:
[[Page 65374]]
1. [Do] the proposed amendment[s] involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
The design function of structures, systems and component[s] are
not Impacted by the proposed change. The proposed change involves
operator manual actions in response to a fire and will not initiate
an event. The proposed actions do not increase the probability of
occurrence of a fire or any other accident previously evaluated.
The proposed actions are feasible and reliable and demonstrate
that the unit can be safely shutdown in the event of a fire. No
significant consequences result from the performance of the proposed
actions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. [Do] the proposed amendment[s] create the possibility of a
new or different kind of accident from any accident previously
evaluated?
Response: No.
The design function of structures, systems and component[s] are
not impacted by the proposed amendment[s]. The proposed change
involves operator manual actions in response to a fire. [It does
not] involve new failure mechanisms or malfunctions that can
initiate a new accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. [Do] the proposed amendment[s] involve a significant
reduction in a margin of safety?
Response: No.
Adequate time is available to perform the proposed operator
manual actions to account for uncertainties in estimates of the time
available and in estimates of how long it takes to diagnose and
execute the actions. The actions are straightforward and do not
create any significant concerns. The actions have been verified that
they can be performed through demonstration and they are
proceduralized. The proposed actions are feasible and reliable and
demonstrate that the unit can be safely shutdown in the event of a
fire.
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 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: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Thomas G. Hiltz.
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.
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 22, 2007.
Brief description of amendment request: The proposed amendment
would allow an alternate methodology from that previously approved in
Topical Report DOM-NAF-3-0.0-P-A, GOTHIC Methodology for Analyzing the
Response to Postulated Pipe Ruptures Inside Containment, as discussed
in the Surry Power Station, Unit Nos. 1 and 2, Updated Final Safety
Analysis Report.
Date of publication of individual notice in Federal Register:
October 30, 2007 (72 FR 61406).
Expiration date of individual notice: Public comment period
expiration date, November 13, 2007; Hearing period expiration date,
January 31, 2008.
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 Systems (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 email to [email protected].
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: December 15, 2006.
Brief description of amendment: The amendment incorporates changes
to the technical specifications (TSs) associated with previously-
approved industry initiatives. The first change relocates the actions
for a safety limit violation from the administrative controls TS
section to the safety limit TS section and deletes notification
requirements, as approved by TS Task Force (TSTF) Change Traveler TSTF-
05-A, ``Deletion of Safety Limit Violation Notification Requirements.''
The second change incorporates generic position titles, as approved by
TSTF-65-A, ``Use of Generic Titles for Utility Positions,'' and
incorporates items approved by Nuclear Regulatory Commission
Administrative Letter 95-06, ``Relocation of Technical
[[Page 65375]]
Specification Administrative Controls Related to Quality Assurance.''
Date of issuance: October 31, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 193.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11386) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket No. 50-413, Catawba Nuclear
Station, Unit 1, York County, South Carolina
Date of application for amendments: November 22, 2006.
Brief description of amendments: The amendment revises the Catawba
Unit 1 Facility Operating License (FOL) to add a license condition
requiring a specific date by which the modifications to the Emergency
Core Cooling Systems (ECCS) sump in response to 2004 Generic Letter
(GL) 2004-02, ``Potential Impact of Debris Blockage on Emergency
Recirculation During Design Basis Accidents at Pressurized Water
Reactors.'' The changes add a license condition which requires that (1)
Catawba Nuclear Station, Unit 1 will enter Mode 5 for the outage to
install the sump strainer modification no later than May 19, 2008, and
that (2) the Unit 1 sump strainer modification will be completed prior
to entry into Mode 4 after May 19, 2008.
Date of issuance: October 31, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 237.
Facility Operating License Nos. NPF-35: Amendment revises the
license.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11386)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: January 4, 2007.
Brief description of amendment: The amendment revised Technical
Specifications (TSs) for the Limiting Conditions for Operation and
Surveillance Requirements for Control Rod Operability, Scram Insertion
Times, and Control Rod Accumulators.
Date of issuance: November 5, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 120 days.
Amendment No.: 230.
Facility Operating License No. DPR-35: The amendment revised the
License and TSs.
Date of initial notice in Federal Register: April 24, 2007 (72 FR
20381).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 5, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353,
Limerick Generating Station, Unit 1 and 2, Montgomery County,
Pennsylvania
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: April 12, 2007.
Brief description of amendments: The amendments modify technical
specification (TS) requirements related to control room envelope
habitability in accordance with TS Task Force (TSTF) Traveler TSTF-448,
Revision 2, ``Control Room Habitability.''
Date of issuance: October 31, 2007.
Effective date: As of the date of issuance, to be implemented
within 180 days.
Amendment Nos.: 150, 150, 145, 145, 178, 186, 173, 188, 149, 264,
and 268.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77,
NPF-62, NPF-11, NPF-18, NPF-39, NPF-85, DPR-44, and DPR-56: The
amendments revised the Technical Specifications and the Operating
Licenses.
Date of initial notice in Federal Register: June 5, 2007 (72 FR
31100).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendment: July 10, 2007.
Brief description of amendment: The amendments revise the value of
the safety limit minimum critical power ratio for the Dresden Nuclear
Power Station (DNPS), Unit 2 technical specifications (TSs). The
amendment also made conforming changes that clarify the wording of the
DNPS, Unit 3 TSs.
Date of issuance: November 6, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 224/216.
Renewed Facility Operating License Nos. DPR-19 and DPR-25: The
amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: July 31, 2007 (72 FR
41783), and September 5, 2007 (72 FR 50986).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 6, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: November 7, 2007, as
supplemented by letter dated January 24, 2007.
Brief description of amendments: The amendments revise Technical
Specification (TS) Surveillance Requirement (SR) 3.4.3.1 to increase
the allowable as-found main steam safety valve lift setpoint tolerance
from 1 percent to 3 percent. In addition, the
amendments revise TS SR 3.1.7.10 to increase the enrichment of sodium
pentaborate used in the standby liquid control system from >=30.0 atom
percent boron-10 to >=45.0 atom percent boron-10.
Date of issuance: November 1, 2007.
Effective date: As of the date of issuance and shall be implemented
prior to main steam safety valve testing during the next refueling
outage currently scheduled for May 2009 for Unit 1 and May 2008 for
Unit 2.
[[Page 65376]]
Amendment Nos.: 235/230.
Renewed Facility Operating License Nos. DPR-29 and DPR-30: The
amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: January 30, 2007 (72 FR
4307) The January 24, 2007, supplement contained clarifying information
and did not change the NRC staff(s initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 1, 2007.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida.
Date of application for amendment: October 11, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.7.7, ``Nuclear Services Closed Cycle Cooling Water
(SW) System,'' to reduce the allowed outage time when one of the
required SW heat exchangers is out of service.
Date of issuance: October 23, 2007.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 225.
Facility Operating License No. DPR-72: Amendment revised the TSs.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6783).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 23, 2007.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: February 8, 2007, as
supplemented by letter dated August 23, 2007.
Brief description of amendment: The amendment changes the basis for
protection of the spent fuel stored in the spent fuel pool (SFP) in
order to eliminate the Final Safety Analysis Report commitment for
maintaining the SFP missile shields.
Date of issuance: October 24, 2007.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 226.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11381). The supplement dated August 23, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 24, 2007.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: October 5, 2006, as supplemented
by letters dated April 4 and July 19, 2007.
Brief description of amendment: The amendment changes the
restrictions on fuel storage in the spent fuel pool.
Date of issuance: October 25, 2007.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 227.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 21, 2006 (71
FR 67394). The supplements dated April 4 and July 19, 2007, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 25, 2007.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: April 22, 2007.
Brief description of amendments: Amendments delete Section 3.H of
Facility Operating License Nos. DPR-67 and NPF-16, which require
reporting of violations of the requirements of Sections 3.A, 3.D, 3.F
and 3.G of the operating license.
Date of Issuance: October 31, 2007.
Effective Date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 203 and 150.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the operating license conditions and Technical
Specifications.
Date of initial notice in Federal Register: June 19, 2007 (72 FR
33783).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: May 4, 2007.
Brief description of amendments: The proposed amendment would
incorporate the administrative changes to Technical Specification (TS)
6.2.1.a, ``On and Offsite Organization'' and 6.8.1.a, ``Procedures and
Programs.''
Date of issuance: November 2, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos: 236 and 231.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TSs.
Date of initial notice in Federal Register: July 3, 2007 (72 FR
36522).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 2, 2007.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2, Oswego County, New York
Date of application for amendment: July 23, 2007.
Brief description of amendment: The amendment modifies Technical
Specification 3.3.2.1, ``Control Rod Block Instrumentation,'' to allow
a new banked position withdrawal sequence for shutdown, using the
Consolidated Line Item Improvement Process.
Date of issuance: October 26, 2007.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 120.
Renewed Facility Operating License No. NPF-69: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: September 25, 2007 (72
FR 54477).
The Commission's related evaluation of the amendment is contained
in a
[[Page 65377]]
Safety Evaluation dated October 26, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant (PINGP), Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: May 10, 2007.
Brief description of amendments: The requested changes are a
partial adoption of Technical Specification Task Force (TSTF)-491,
Revision 2, ``Removal of Main Steam and Feedwater Valve Isolation
Times'' which was proposed by the TSTF by letter on May 18, 2006. The
proposed changes revise Technical Specification (TS) 3.7.2 ``Main Steam
Valves Closure Times'' by relocating the isolation valve closure times
to a licensee-controlled document identified as a Bases reference. The
proposed amendments deviate from TSTF-491 in that the current PINGP TS
(3.7.3) and associated surveillance requirements for the main feedwater
isolation valves do not include valve closure times, and thus, the
changes to TS 3.7.3 provided for in TSTF-491 are not applicable to the
PINGP TSs and are not adopted. TSTF change traveler TSTF-491, Revision
2, was announced for availability in the Federal Register on December
29, 2006, as part of the consolidated line item improvement process.
Date of issuance: October 31, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 181 and 171.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 17, 2007 (72 FR
39083).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California.
Date of application for amendments: April 17, 2007.
Brief description of amendments: The amendment modified Technical
Specifications requirements related to control room envelope
habitability in accordance with Technical Specifications Task Force
448, Revision 3, using the Consolidated Line Item Improvement Process.
Date of issuance: October 31, 2007.
Effective date: as of its date of issuance, to be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2-214; Unit 3-206.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 22, 2007 (72 FR
28722). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: May 21, 2007, as supplemented by
letter dated June 11, 2007.
Brief description of amendment: The amendment modified the
technical specification (TS) requirements for inoperable snubbers by
adding Limited Condition for Operation 3.0.8, using the Consolidated
Line Item Improvement Process. The change is based on TS Task Force
(TSTF) TSTF-372, Revision 4.
Date of issuance: October 17, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 251, 231.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the licenses and the technical specifications.
Date of initial notice in Federal Register: June 19, 2007 (72 FR
33785) The supplement dated July 11, 2007, 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 October 17, 2007.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: May 29, 2007.
Brief description of amendment: The amendments modify the Technical
Specification requirements related to control room habitability, using
the Technical Specification Task Force traveler, TSTF-448, revision 3.
Date of issuance: October 31, 2007.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment Nos.: 252, 232.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the licenses and the technical specifications.
Date of initial notice in Federal Register: July 3, 2007 (72 FR
36523).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 8th day of November 2007.
For The Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-22331 Filed 11-19-07; 8:45 am]
BILLING CODE 7590-01-P