[Federal Register Volume 73, Number 38 (Tuesday, February 26, 2008)]
[Notices]
[Pages 10293-10302]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-3481]
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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 to 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 January 31 to February 13, 2008. The last
biweekly notice was published on February 12, 2008 (73 FR 8068).
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.
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-
[[Page 10294]]
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
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 support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to 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 ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM is free and is
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html.
[[Page 10295]]
Information about applying for a digital ID certificate is available on
NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically 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, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. 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.
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].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: August 13, 2007.
Description of amendments request: The amendment would revise
Technical Specification (TS) Table 3.3.1.2-1, ``Source Range Monitor
[SRM] Instrumentation,'' to add a note that specifies the required
locations of SRMs in Mode 5 during core alterations, and also to make
an administrative correction to Unit 1 TS Surveillance Requirement (SR)
3.3.1.2.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are administrative in nature. There are no
requirements being added, deleted, or altered as a result of either
of the proposed changes.
The change to Table 3.3.1.2-1 adds a footnote to Table 3.3.1.2-1
which duplicates the Mode 5 operable SRM location requirements
currently specified in SR 3.3.1.2.2 and discussed in the LCO
[limiting condition for operation] bases section for TS 3.3.1.2. The
specific Mode 5 operable SRM location requirements are not being
changed and are consistent with the requirements provided in the
current version of NUREG-1433. This change is being done as an aid
to Operations personnel, to help prevent inadvertently missing the
requirements.
The change to SR 3.3.1.2.2 for Unit 1 corrects a typographical
error to be consistent with other locations within the Unit 1 and
Unit 2 TSs as well as the current version of NUREG 1433.
The proposed changes do not involve a physical change to the
SRMs, nor do they alter the assumptions of the accident analyses.
Therefore, the probability and the consequences of an accident
previously evaluated are not affected.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical change to the
SRMs, nor do they alter the assumptions of the accident analyses.
The changes are purely administrative in nature. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are administrative in nature, being done as
an aid to Operations personnel, to help prevent inadvertently
missing the Mode 5 operable SRM location requirements and to correct
a typographical error. There are no requirements being added,
deleted, or altered as a result of either
[[Page 10296]]
of the proposed changes. As such, the proposed changes do not
involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: January 15, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) Surveillance Requirement (SR)
frequency in TS 3.1.3, ``Control Rod OPERABILITY'' from ``7 days after
the control rod is withdrawn and THERMAL POWER is greater than the [Low
Power Setpoint] LPSP of [Rod Worth Minimizer] RWM'' to ``31 days after
the control rod is withdrawn and THERMAL POWER is greater than the LPSP
of the RWM'' and revise Example 1.4-3 in Section 1.4 ``Frequency'' to
clarify the applicability of the 1.25 surveillance test interval
extension. The proposed amendment does not adopt the clarification of
Source Range Monitor (SRM) TS action for inserting control rods, which
is applicable only to Boiling Water Reactor (BWR)/6 plants. Since Fermi
2 is a BWR/4 plant, this change in TSTF-475, Revision 1 is not
applicable and therefore, not adopted.
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 by a reference to a generic analysis published in the
Federal Register on November 13, 2007 (72 FR 63935), which is presented
below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated Biweekly Notice Coordinator.
The proposed change generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM Insert Control Rod
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and
NUREG-1434 (BWR/6) STS. The changes: (1) Revise TS testing frequency
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod
OPERABILITY'', [ ], and (3) revise Example 1.4-3 in Section 1.4
``Frequency'' to clarify the applicability of the 1.25 surveillance
test interval extension. The consequences of an accident after
adopting TSTF-475, Revision 1 are no different than the consequences
of an accident prior to adoption. 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 Accident Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change 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 analyzed. Thus, 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.
TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2
frequency in TS 3.1.3, ``Control Rod OPERABILITY'', [ ], and (3)
revise Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the
applicability of the 1.25 surveillance test interval extension. The
GE Nuclear Energy Report, ``CRD Notching Surveillance Testing for
Limerick Generating Station,'' dated November 2006, concludes that
extending the control rod notch test interval from weekly to monthly
is not expected to impact the reliability of the scram system and
that the analysis supports the decision to change the surveillance
frequency. Therefore, the proposed changes in TSTF-475, Revision 1 [
] 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: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Acting Branch Chief: Patrick Milano.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: July 30, 2007.
Description of amendment request: The amendments would revise the
Technical Specifications to allow single header operation of the
nuclear service water system (NSWS) for a time period of 35 days. The
change will facilitate future maintenance of the NSWS headers.
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:
[First Standard]
Does operation of the facility in accordance with the proposed
amendment involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed single supply header operation configuration for
NSWS operation and the associated proposed TS and Bases changes have
been evaluated to assess their impact on plant operation and to
ensure that the design basis safety functions of safety related
systems are not adversely impacted. During single supply header
operation, the operating NSWS header will be able to supply all
required NSWS flow to safety related components. It was demonstrated
that proposed single failures would not cause the NSWS to be
rendered incapable of performing its required safety related
function under accident conditions.
The purpose of this amendment request is to ultimately
facilitate inspection and maintenance of the NSWS supply headers.
Therefore, NRC approval of this request will ultimately help to
enhance the long-term structural integrity of the NSWS and will help
to ensure the system's reliability for many years.
In general, the NSWS serves as an accident mitigation system and
cannot by itself initiate an accident or transient situation. The
only exception is that the NSWS piping can serve as a source of
floodwater to safety related equipment in the auxiliary building or
in the diesel generator buildings in the event of a leak or a break
in the system piping. The probability of such an event is not
significantly increased as a result of this proposed request. NSWS
piping added in support of the proposed request will be tested and
maintained in a manner consistent with that for comparable safety
related piping in the NSWS.
The proposed 35 day TS Required Action Completion Time has been
evaluated for risk significance and the results of this evaluation
have been found acceptable. The probabilities of occurrence of
accidents presented in the UFSAR will not increase as a result of
implementation of this change. Because the PRA analysis supporting
the proposed change yielded acceptable results, the NSWS will
maintain its required availability in response to accident
situations. Since NSWS availability is maintained, the response of
the plant to accident situations will remain acceptable and the
consequences of accidents presented in the UFSAR will not increase.
[Second Standard]
Does operation of the facility in accordance with the proposed
amendment create the
[[Page 10297]]
possibility of a new or different kind of accident from any accident
previously evaluated?
Response: No.
Implementation of this amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed request does not affect the basic operation
of the NSWS or any of the systems that it supports. These include
the Emergency Core Cooling System, the Containment Spray System, the
Containment Valve Injection Water System, the Auxiliary Feedwater
System, the Component Cooling Water System, the Control Room Area
Ventilation System, the Control Room Area Chilled Water System, the
Auxiliary Building Filtered Ventilation Exhaust System, or the
Diesel Generators. During proposed single supply header operation,
the NSWS will remain capable of fulfilling all of its design basis
requirements, even when assuming the required single failure.
No new accident causal mechanisms are created as a result of NRC
approval of this amendment request. No changes are being made to the
plant which will introduce any new type of accident outside those
assumed in the UFSAR.
[Third Standard]
Does operation of the facility in accordance with the proposed
amendment involve a significant reduction in the margin of safety?
Response: No.
Implementation of this amendment will not involve a significant
reduction in any margin of safety. Margin of safety is related to
the confidence in the ability of the fission product barriers to
perform their design functions during and following an accident
situation. These barriers include the fuel cladding, the reactor
coolant system, and the containment system. The performance of these
fission product barriers will not be impacted by implementation of
this proposed TS amendment. During single supply header operation,
the NSWS and its supported systems will remain capable of performing
their required functions even assuming the postulated single
failure. No safety margins will be impacted.
The PRA conducted for this proposed amendment demonstrated that
the impact on overall plant risk remains acceptable during single
supply header operation. Therefore, there is not 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: 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: Melanie C. Wong, Acting.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 27, 2007.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.7.2 (Main Steam Isolation Valves) and TS
3.7.3 (Main Feedwater Isolation Valves, Main Feedwater Control Valves,
Associated Bypass Valves and Tempering Valves) by removing the specific
isolation time for the isolation valves from the associated
Surveillance 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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
Criterion 1: The Proposed Changes Do Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed changes allow relocating main steam and main feedwater
valve isolation times to the licensee-controlled document that is
referenced in the Bases. The proposed changes are described in
Technical Specification Task Force (TSTF) Standard TS Change Traveler
TSTF-491 related to relocating the main steam and main feedwater valves
isolation times to the licensee-controlled document that is referenced
in the Bases and replacing the isolation time with the phrase, ``within
limits.'' The proposed changes do not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
The proposed changes relocate the main steam and main feedwater
isolation valve times to the licensee-controlled document that is
referenced in the Bases. The requirements to perform the testing of
these isolation valves are retained in the TSs. Future changes to the
Bases or licensee-controlled document will be evaluated pursuant to the
requirements of 10 CFR 50.59, ``Changes, test and experiments,'' to
ensure that such changes do not result in more than a minimal increase
in the probability or consequences of an accident previously evaluated.
The proposed changes do not adversely affect accident initiators or
precursors nor alter the design assumptions, conditions, and
configuration of the facility 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 (SSCs) 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, or
radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and the amounts
of radioactive effluent that may be released, nor significantly
increase individual or cumulative occupational/public radiation
exposures. Therefore, the changes do not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2: The Proposed Changes Do Not Create the Possibility of
a New or Different Kind of Accident from any Previously Evaluated.
The proposed changes relocate the main steam and main feedwater
valve isolation times to the licensee-controlled document that is
referenced in the Bases. In addition, the valve isolation times are
replaced in the TS with the phrase ``within limits''. The 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 TSs continue to require
testing of the main steam and main feedwater isolation valves to ensure
the proper functioning of these isolation valves. Therefore, the
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
Criterion 3: The Proposed Changes Do Not Involve a Significant
Reduction in the Margin of Safety.
The proposed changes relocate the main steam and main feedwater
valve isolation times to the licensee-controlled document that is
referenced in the Bases. In addition, the valve isolation times are
replaced in the TSs with the phrase ``within limits.'' Instituting the
proposed changes will continue to ensure the testing of main steam and
main feedwater isolation valves. Changes to the Bases or license-
controlled document are performed in accordance with 10 CFR 50.59. This
approach provides an effective level of regulatory control and ensures
that main steam and feedwater isolation valve testing is conducted such
that there is no significant reduction in the margin of safety. The
margin of safety provided by the isolation valves is unaffected by the
proposed changes since there continue to be TS requirements to ensure
the testing of main steam and main
[[Page 10298]]
feedwater isolation valves. The proposed changes maintain sufficient
controls to preserve the current margins of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: 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: Melanie C. Wong, Acting.
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 amendment request: December 21, 2007.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) Surveillance Requirements (SR) 3.8.4.2 and
3.8.4.5 to add an additional acceptance criterion to verify that total
battery connector resistance is within pre-established limits that
ensure the batteries can perform their design functions. The proposed
amendment is in response to a non-cited violation that was documented
in NRC Component Design Bases Inspection Report 05000254/2006003(DRS),
05000265/2006003(DRS).
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 revisions of SR 3.8.4.2 and SR 3.8.4.5 to add a battery
connector resistance acceptance criterion will not challenge the
ability of the safety-related batteries to perform their safety
function. Appropriate monitoring and maintenance will continue to be
performed on the safety-related batteries. In addition, the safety-
related batteries are within the scope of 10 CFR 50.65,
``Requirements for monitoring the effectiveness of maintenance at
nuclear power plants,'' which will ensure the control of maintenance
activities associated with this equipment.
Current TS requirements will not be altered and will continue to
require that the equipment be regularly monitored and tested. Since
the proposed change does not alter the manner in which the batteries
are operated, there is no significant impact on reactor operation.
The proposed change does not involve a physical change to the
batteries, nor does it change the safety function of the batteries.
The proposed TS revision involves no significant changes to the
operation of any systems or components in normal or accident
operating conditions and no changes to existing structures, systems,
or components.
Therefore, these changes will not increase 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 revising SR 3.8.4.2 and SR 3.8.4.5 to add
an additional acceptance criterion for battery connector resistance
is an increase in conservatism, without a change in system testing
methods, operation, or control. Safety-related batteries installed
in the plant will be required to meet criteria more restrictive and
conservative than current acceptance criteria and standards. The
proposed change does not affect the manner in which the batteries
are tested and maintained; therefore, there are no new failure
mechanisms for the system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the setpoints for the actuation of
equipment relied upon to respond to an event. The proposed change
does not modify the safety limits or setpoints at which protective
actions are initiated. The change is conservative and further
ensures safety-related battery operability and availability.
As such, sufficient DC capacity to support operation of
mitigation equipment is enhanced, which results in an increase in
the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: December 20, 2007.
Description of amendment request: Duane Arnold Energy Center (DAEC)
requests a change, consistent with the adoption of TSTF-475, Revision
1, an approved change to the Standard Technical Specifications (STS)
for General Electric (GE) Plants (NUREG-1433, BWR/4) and plant specific
technical specifications (TS), that allows: (1) Revising the frequency
of Surveillance Requirement (SR) 3.1.3.2, notch testing of fully
withdrawn control rod, from ``7 days after the control rod is withdrawn
and THERMAL POWER is greater than 20% [Rated Thermal Power] RTP'' to
``31 days after the control rod is withdrawn and THERMAL POWER is
greater than 20% RTP'' and (2) revising Example 1.4-3 in Section 1.4
``Frequency'' to clarify that the 1.25 surveillance test interval
extension in SR 3.0.2 is applicable to time periods discussed in NOTES
in the ``SURVEILLANCE'' column in addition to the time periods in the
``FREQUENCY'' column.
The NRC staff acknowledges that, in item (1) above, the wording
that is to be adopted by the Duane Arnold TS in SR 3.1.3.2 (``31 days
after the control rod is withdrawn and THERMAL POWER is greater than
20% RTP'') is a deviation from the language in the Improved STS (``31
days after the control rod is withdrawn and THERMAL POWER is greater
than the [Low Power Setpoint] LPSP of the [Rod Worth Minimizer] RWM.'')
This deviation from NUREG-1433 was incorporated into the DAEC TS by
Amendment 223 dated May 22, 1998, in the conversion of the DAEC TS to
the Improved STS.
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 (NSHC) through incorporation by reference of the NSHC
determination (NSHCD) published in the Federal Register Notice dated
November 13, 2007, that announced the availability of TS improvement
through the consolidated line item improvement process (CLIIP). The
NSHCD, with references to BWR/6 information deleted and with clarifying
comments inserted within brackets [ ], 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 generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM Insert Control Rod
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4)
[[Page 10299]]
STS. The changes: (1) Revise TS testing frequency for surveillance
requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod OPERABILITY''
and (2) revise Example 1.4-3 in Section 1.4 ``Frequency'' to clarify
the applicability of the 1.25 surveillance test interval extension.
The consequences of an accident after adopting TSTF-475,
Revision 1 are no different than the consequences of an accident
prior to adoption. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change 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 analyzed. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
TSTF-475, Revision 1 [, as adopted by DAEC TS,] will: (1) Revise
the TS SR 3.1.3.2 frequency in TS 3.1.3, ``Control Rod OPERABILITY''
and (2) revise Example 1.4-3 in Section 1.4 ``Frequency'' to clarify
the applicability of the 1.25 surveillance test interval extension.
The GE Nuclear Energy Report, ``CRD Notching Surveillance
Testing for Limerick Generating Station,'' dated November 2006,
concludes that extending the control rod notch test interval from
weekly to monthly is not expected to impact the reliability of the
scram system and that the analysis supports the decision to change
the surveillance frequency. Therefore, the proposed changes in TSTF-
475, Revision 1 [, as adopted by DAEC TS,] 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: Marjan Mashhadi, Florida Power & Light
Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.
NRC Acting Branch Chief: Patrick Milano.
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 e-mail to [email protected].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: November 15, 2007, as
supplemented by letter dated December 21, 2007.
Brief description of amendment: The amendment is a one-time change
that revised Technical Specification (TS) Section 3.1.7, ``Rod Position
Indication.'' The requirements related to one inoperable bank demand
position indicator (DPI) are modified by a footnote to allow two DPIs
to be inoperable per bank for one or more banks on a temporary basis
during the current operating cycle (Cycle 25). This provision allows
for corrective maintenance on three inoperable DPIs in the rod position
indication system that necessitates removing both DPIs for the affected
rod banks from service during the repair. This amendment expires at the
end of operating Cycle 25.
Date of issuance: January 29, 2008.
Effective date: Effective as of the date of issuance and shall be
implemented within 60 days.
Amendment No. 217.
Renewed Facility Operating License No. DPR-23: The amendment
revises the Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: November 28, 2007 (72
FR 67321).
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment and final NSHC
determination are contained in a safety evaluation dated January 29,
2008.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602-1551.
NRC Branch Chief: Thomas H. Boyce.
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: October 2, 2007.
Brief description of amendment: The amendment revises Technical
Specification 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 amendment adds a new specification when two EDGs are
inoperable and revises the surveillance requirements for the EDGs.
Date of issuance: February 7, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 194.
Facility Operating License No. DPR-43: Amendment revised the
License and Technical Specifications.
[[Page 10300]]
Date of initial notice in Federal Register: November 20, 2007 (72
FR 65363)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 7, 2008.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: February 16, 2007.
Brief description of amendment: The proposed amendment would revise
Technical Specification 3/4.4.3, ``Reactor Coolant System, Relief
Valves'' to modify the method of testing the pressurizer Power Operated
Relief Valves (PORVs). Specifically, the requirement for bench testing
the valves is changed to accommodate testing of the PORVs while
installed in the plant. The change is requested due to the installation
of new PORVs that are welded to the piping rather than bolted into the
system.
Date of issuance: February 12, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 302.
Facility Operating License No. DPR-65: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: November 19, 2007 (72
FR 65084).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 12, 2008.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Date of amendment request: April 24, 2007, as supplemented by
letter dated August 2, 2007, and electronic mail dated January 8, 2008.
Brief description of amendments: The amendments relocate the Fuel
Handling Area Ventilation System and associated Ventilation Filter
Testing Program requirements that are included in the Unit 1 Technical
Specifications (TS) 3.7.12 and 5.5.11 and the Unit 2 TS 3.9.11 and
6.5.11 to the unit-specific Technical Requirements Manuals (TRMs). The
TRMs are licensee-controlled documents which are controlled under 10
CFR 50.59, ``Changes, tests, and experiments.''
Date of issuance: February 4, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1-231; Unit 2-274.
Renewed Facility Operating License Nos. DPR-51 and NPF-6:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 5, 2007 (72 FR
31098). The supplemental letter dated August 2, 2007, and electronic
mail dated January 8, 2008, 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 amendments
is contained in a Safety Evaluation dated February 4, 2008.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: October 24, 2007.
Brief description of amendment: The amendment revises the
containment buffering agent used for pH control under post loss-of-
coolant accident (LOCA) conditions, from trisodium phosphate to sodium
tetraborate.
Date of issuance: February 7, 2008.
Effective date: As of the date of issuance, and shall be
implemented prior to entry into Mode 4 following completion of the
spring 2008 refueling outage.
Amendment No.: 253.
Facility Operating License Nos. DPR-26: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: December 4, 2007 (72 FR
68211).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 7, 2008.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: July 25, 2007, as supplemented
November 1, 2007.
Brief description of amendment: The proposed amendment would modify
the Technical Specifications by adding an Action Statement to the
Limiting Conditions for Operation (LCOs) for TS 3.7.4, ``Control Room
Air Conditioning (AC) System.'' Specifically, the new Action statement
allows 72 hours to restore one control room air conditioning subsystem
to operable status and requires verification that the control room
temperature remains below 90 [deg]F every 4 hours during the period of
inoperability. The change is consistent with NRC-approved Revision 3 to
Technical Specifications Task Force (TSTF) Improved Standard Technical
Specifications Change Traveler, TSTF-477, ``Add Action Statement for
Two Inoperable Control Room Air Conditioning Subsystems.''
Date of issuance: January 23, 2008.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 290.
Facility Operating License No. DPR-59: The amendment revises the
License and the Technical Specifications.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51855).
The November 1, 2007, supplement provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 23, 2008.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: November 6, 2006, supplemented
by letters dated August 10, 2007, and December 20, 2007.
Brief description of amendment: The amendment would revise Appendix
A, technical specification (TS), Core Operating Limits Report
analytical methods referenced in TS 5.6.5.b to add EMF-2103 (P)(A),
``Realistic Large Break LOCA Methodology for Pressurized Water
Reactors.''
Date of issuance: January 31, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 229.
Facility Operating License No. DPR-20: Amendment revised the
technical specifications.
[[Page 10301]]
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75995)
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendments
is contained in a Safety Evaluation dated January 31, 2008.
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
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1
and 2, Will County, Illinois.
Date of application for amendment: January 8, 2007 as supplemented
by letter dated October 12, 2007.
Brief description of amendment: The amendments extended the reactor
trip system and engineered safety features actuation system completion
times, bypass test times, and surveillance test intervals for technical
specifications (TS) 3.3.1, ``RTS Instrumentation,'' TS 3.3.2, ``ESFAS
Instrumentation,'' and TS 3.3.6, ``Containment Ventilation Isolation
Instrumentation.''
Date of issuance: January 29, 2008.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 153, 153, 148, and 148.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: March 27, 2007 (72 FR
14305).
The October 12, 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 January 29, 2008.
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: November 12, 2007.
Brief description of amendments: The amendments revise TS 3.1.3.2,
``Position Indication Systems--Operating,'' to allow for the use of an
alternate method, other than the movable incore detectors, to monitor
the position of a control rod or shutdown rod in the event of a problem
with the analog rod position indication system. The use of this
alternate method will reduce the required frequency of flux mapping
using the movable incore detectors to determine the position of the
non-indicating rod, thus reducing the wear on the movable incore
detector system that is also used to complete other required TS
surveillances.
Date of issuance: January 28, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos: 237 and 232.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 28, 2007 (72
FR 67323).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 28, 2008.
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: September 19, 2007.
Brief description of amendment: The amendment revises Limiting
Condition for Operation 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, using the Consolidated Line Item
Improvement Process.
Date of issuance: February 7, 2008.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 121.
Renewed Facility Operating License No. NPF-69: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: November 20, 2007 (72
FR 65368).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 7, 2008.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: February 15, 2007, as
supplemented on November 30, 2007.
Brief description of amendment: The amendment revised the Technical
Specifications Surveillance Requirement (SR) 3.8.4.2, ``DC [Direct
Current] Sources--Operating,'' to specify that the Division 1 battery
chargers are verified to supply >=150 amps and the Division 2 battery
chargers are verified to supply >=110 amps.
Date of issuance: January 30, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 153.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 24, 2007 (72 FR
20384).
The supplemental letter contained clarifying information, did not
change the initial no significant hazards consideration determination,
and did not expand the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated January 30, 2008.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 21, 2007.
Brief description of amendment: The amendment revises Technical
Specifications (TS) safety limit (SL) requirements related to the use
of a non-cycle specific peak linear heat rate (PLHR) SL of 22 kW/ft to
fuel centerline melt (FCM). The TS change is consistent with the
Technical Specification Task Force (TSTF) 445-A, Revision 1. Because
these Limiting Safety Systems Setting (LSSS) values appear in the FCS
TS Bases Sections of TS 1.3, TS 1.0, Safety Limits and Limiting Safety
System Settings, was also revised to more clearly align with the
Combustion Engineering (CE) Standard Technical Specifications (STS) 2.0
in content. Therefore, TS Section 1.1, Safety Limits--Reactor Core, is
revised to incorporate the TSTF-445-A, Revision 1, peak fuel centerline
temperature criteria and TS 1.2, Safety Limits--Reactor Coolant System
Pressure, is revised to incorporate the SL violation
[[Page 10302]]
action which is currently delineated in administrative control TS
5.7.1. TS Section 1.3, Limiting Safety System Settings, was relocated
to the currently unused TS Section 2.13 to be more consistent with the
content of the CE STS (i.e., the LSSS will be located in the Limiting
Conditions for Operation (LCO) section of the FCS TS which is similar
to the LCO/Surveillance Requirements Section 3.0 of the STS). As noted
above, the administrative control in TS 5.7.1, Safety Limit Violation,
is relocated. Also, administrative control TS 5.9.5, Core Operating
Limits Report (COLR), item a., is revised to add TS 2.13, RPS Limiting
Safety System Settings, Table 2-11, Items 6, 8, and 9, to the list of
items that shall be documented in the COLR. The TS Table of Contents
(TOC) is also updated to reflect the deletion and subsequent
renumbering of Section 1.3 and Table 1-1 to TS 2.13 and Table 2-11,
respectively. The TOC is also updated to delineate the new TS
subsections 1.1.1 and 1.1.2, provide the revised titles for TS 1.0,
1.1, 1.2, and 2.13, and to reflect TS 5.7.1 as ``Not used.''
Date of issuance: February 4, 2008.
Effective date: As of its date of issuance and prior to startup
from the 2008 refueling outage.
Amendment No.: 252.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: November 6, 2007 (72 FR
62690). The Commission's related evaluation of the amendment is
contained in a safety evaluation dated February 4, 2008.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: October 11, 2007, as
supplemented on October 25, December 4 and 26, 2006, February 13, March
14 and 22, April 13, 17, 23, 26, and 27, May 3, 9, 14, and 21, June 1,
4, 8, 14, 20, and 27, July 6, 12, 13, 30, and 31, August 3, 13, 15, and
28, September 19, October 5, November 30, December 10, 2007, and
January 9, 24, and 29, 2008.
Brief description of amendments: The amendments increase the SSES 1
and 2 licensed thermal power to 3952 Mega-watts thermal (MWt), which is
20% above the original rated thermal power (RTP) of 3293 MWt, and
approximately 13% above the current RTP of 3489 MWt. The amendments
revise the SSES 1 and 2 Operating License and Technical Specifications
necessary to implement the increased power level.
Date of issuance: January 30, 2008.
Effective date: As of the date of issuance and to be implemented in
accordance with the issued License Conditions.
Amendment Nos.: 246 and 224.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11392). The supplements dated October 25, December 4 and 26, 2006,
February 13, March 14 and 22, April 13, 17, 23, 26, and 27, May 3, 9,
14, and 21, June 1, 4, 8, 14, 20, and 27, July 6, 12, 13, 30, and 31,
August 3, 13, 15, and 28, September 19, October 5, November 30,
December 10, 2007, and January 9, 24, and 29, 2008, 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 January 30, 2008.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 15th day of February 2008.
For The Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E8-3481 Filed 2-25-08; 8:45 am]
BILLING CODE 7590-01-P