[Federal Register Volume 74, Number 75 (Tuesday, April 21, 2009)]
[Notices]
[Pages 18251-18262]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-8832]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0170]
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 March 26, 2009, to April 8, 2009. The last
biweekly notice was published on April 7, 2009 (74 FR 15765).
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-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking
and Directives Branch, TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. 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.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area 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
[[Page 18252]]
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.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve all adjudicatory documents
over the Internet or in some cases to mail copies on electronic storage
media. Participants may not submit paper copies of their filings unless
they seek 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. 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 electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday, excluding government holidays. The help
electronic filing Help Desk can be contacted by telephone at 1-866-672-
7640 or by e-mail at [email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file 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).
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings, unless an NRC regulation or
other law requires submission of such information. With respect to
copyrighted works, except for limited excerpts that serve the purpose
of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their 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-
[[Page 18253]]
4209, (301) 415-4737 or by e-mail to [email protected].
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: August 21, 2008.
Description of amendment request: The proposed amendments would
revise the proposed license amendment implements Technical
Specification Task Force (TSTF) Changes Travelers TSTF-479, Revision 0,
``Changes to Reflect Revision of [Title 10 of the Code of Federal
Regulations] 10 CFR 50.55a'' and TSTF-497, Revision 0, ``Limit
Inservice Testing [IST] Program SR 3.0.2 Application to Frequencies of
2 Years or Less''. TSTF-479 and TSTF-497 revise the technical
specification Administrative Controls section pertaining to
requirements for the IST Program, consistent with the requirements of
10 CFR 50.55a(f)(4) for pumps and valves which are classified as
American Society of Mechanical Engineers (ASME) Code Class 1, Class 2,
and Class 3.
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 revises TS [Technical Specification] 5.5.8,
``Inservice Testing Program,'' for consistency with the requirements
of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and
valves which are classified as ASME Code Class 1, Class 2, and Class
3. The proposed change incorporates revisions to the ASME [American
Society of Mechanical Engineers] Code as identified in the TSTFs
[Technical Specification Task Force] referenced above.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. The proposed change does not involve the addition or removal
of any equipment, or any design changes to the facility.
Additionally, there is no change in the types or increases in the
amounts of any effluent that may be released offsite and there is no
increase in individual or cumulative occupational exposure.
Therefore, this 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 revises TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves
which are classified as ASME Code Class 1, Class 2, and Class 3. The
proposed change incorporates revisions to the ASME Code as
identified in the TSTFs referenced above. The proposed change does
not involve a modification to the physical configuration of the
plant nor does it involve a change in the methods governing normal
plant operation. The proposed change will not impose any new or
different requirements or introduce a new accident initiator,
accident precursor, or malfunction mechanism. Additionally, there is
no change in the types or increases in the amounts of any effluent
that may be released offsite and there is no increase in individual
or cumulative occupational exposure.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No
The proposed change revises TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves
which are classified as ASME Code Class 1, Class 2, and Class 3. The
proposed change does not involve a modification to the physical
configuration of the plant nor does it change the methods
governingnormal plant operation. The proposed change incorporates
revisions to the ASME Code as identified in the TSTFs referenced
above.
The safety function of the affected pumps and valves will be
maintained.
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: Melanie Wong.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: February 16, 2009.
Description of amendment request: The Arkansas Nuclear One, Unit
No. 1 (ANO-1) Technical Specification (TS) 5.5.16, ``Reactor Building
Leakage Rate Testing Program,'' contains reactor building leak rate
criteria for overall Type A, B, and C testing. However, TS 5.5.16 does
not specify criteria for Type B air lock leakage testing. Entergy
Operations, Inc., proposes to modify TS 5.5.16 to add criteria for
overall air lock leakage testing and to adopt a low pressure test
method relevant to the air lock door seals. This change is consistent
with NUREG 1430, Revision 3.1, ``Standard Technical Specifications
(STS) for Babcock & Wilcox Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The reactor building air locks are passive components integral
to the reactor building structure and are not associated with
accident initiators. Each air lock door is rated for and tested to
the maximum calculated post-accident pressure of the reactor
building. The air lock door seal pressure test is performed any time
the air lock is used for reactor building access during modes of
operation when reactor building integrity is required and prior to
establishing reactor building integrity. The door seal test is
intended to be a gross test to verify that the door seals were not
damaged during the opening and closing cycle(s). This test does not
replace the required overall barrel leakage test. Based on
information provided by the air lock vendor, a test pressure of 10
psig [pounds per square inch gauge] is conservatively sufficient to
perform this gross seal verification. This new acceptable leakage
rate and test criteria are consistent with NUREG 1430, Rev. 3.1,
Standard Technical Specifications for Babcock & Wilcox Plants (STS)
and are applicable to ANO-1. While new to the TSs, the ANO-1 program
for ensuring compliance with 10 CFR 50, Appendix J has verified
leakage within the proposed limiting values.
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.
No physical changes to the facility are initiated by the
proposed change. In addition, the proposed change has no affect on
plant configuration, or method of operation of plant structures,
systems, or components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not increase the allowable overall air
lock leakage rate, nor
[[Page 18254]]
affect the acceptance criteria of the overall integrated containment
leakage rate as currently tested to in accordance with the ANO-1
containment leakage rate test program. All of the changes are
bounded by existing analyses for all evaluated accidents and do not
create any situations that alter the assumptions used in these
analyses.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: March 10, 2009.
Description of amendment request: The proposed amendment consists
of changes to Technical Specification (TS) 3.4.9, ``Pressurizer,''
which contains a maximum and minimum level for the pressurizer. The
licensee proposes to delete the minimum level requirement. This change
is consistent with NUREG 1430, Rev. 3.1, ``Standard Technical
Specifications [STS] for Babcock and Wilcox Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The minimum Pressure level limit currently specified in the TSs
does not act to ensure specified fuel design limits are protected.
Accident and transient analyses assume lowering or a loss of
Pressurizer level. Safety systems are designed and maintained
available to mitigate the consequences of an accident or transient
that may involve a loss of Reactor Coolant System (RCS) inventory.
None of these systems rely upon a predetermined minimum Pressurizer
level in order to perform their intended function. Furthermore, the
minimum Pressure level limit is unrelated to any anticipated
accident initiator.
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.
No physical changes to the facility are initiated by the
proposed change. In addition, the proposed change has no affect on
plant configuration, or method of operation of plant structures,
systems, or components.
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.
Installed automatic control systems will continue to maintain
Pressurizer level at a predetermined setpoint and are independent of
a prescribed minimum TS level limit. The deletion of the current TS
limit has no impact on guidance or operational response to
pressurizer level deviations. Furthermore, the minimum Pressure
level limit is not an assumed value for accident prevention or
mitigation in the [Arkansas Nuclear One, Unit 1] [Safety Analysis
Report].
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: February 25, 2009.
Description of amendment request: The proposed change removes the
reactor coolant system (RCS) structural integrity requirements
contained in Technical Specification (TS) 3/4.4.8, which specifies
requirements relating to the structural integrity of American Society
of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components. This
specification is redundant to the requirements contained within Title
10 of the Code of Federal Regulations (10 CFR) section 50.55a, ``Codes
and standards.'' With this proposed change, RCS pressure boundary
structural integrity will continue to be maintained by compliance with
10 CFR 50.55a, as implemented through the Limerick Generating Station,
Units 1 and 2, Inservice Inspection 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, with NRC edits in brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to remove the RCS structural integrity
controls from the TSs does not impact any mitigation equipment or
the ability of the RCS pressure boundary to fulfill any required
safety function. Since no accident mitigation [equipment] or
initiators are impacted by this change, no design basis accidents
are affected. Therefore, the proposed change does 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 proposed change will not alter the plant configuration or
change the manner in which the plant is operated. No new failure
modes are being introduced by the proposed change.
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.
Removal of TS 3/4.4.8 from the TSs does not reduce the controls
that are required to maintain the RCS pressure boundary for ASME
Code Class 1, 2, or 3 components.
No equipment or RCS safety margins are impacted due to the
proposed change. 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: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
[[Page 18255]]
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 amendment request: February 16, 2009.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) by removing the structural
integrity requirements contained in TS 3/4.4.10 and the associated TS
bases from the TSs. Removal of TS 3/4.4.10 is consistent with NUREG-
1431, Revision 3.0, ``Standard Technical Specifications Westinghouse
Plants,'' in that it does not meet the criteria of Title 10 of the Code
of Federal Regulations (10 CFR), Part 50, Section 50.36, ``Technical
Specifications,'' for inclusion in the TSs. The proposed amendment
would also relocate the reactor coolant pump (RCP) flywheel inspection
requirements in Surveillance Requirement (SR) 4.4.10 to SR 4.0.5, and
would revise the RCP flywheel inspection interval from 10 years to 20
years. The RCP flywheel inspection interval change is consistent with
Nuclear Regulatory Commission approved Industry/Technical Specification
Task Force (TSTF) Standard Technical Specification Change Traveler,
TSTF-421, ``Revision to RCP Flywheel Inspection Program (WCAP-15666).''
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 remove structural integrity controls from
the TSs does not impact any mitigation equipment or the ability of
the RCS [reactor coolant system] pressure boundary to fulfill any
required safety function. The proposed change will continue to
ensure the requirements of 10 CFR 50.55a [``Codes and standards'']
are maintained as specified in TS 4.0.5. Since no accident
mitigation or initiators are impacted by this change, no design
basis accidents are affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any 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 change will not alter the plant configuration or
change the manner in which the plant is operated. Structural
integrity will continue to be maintained as required by 10 CFR
50.55a and specified in TS 4.0.5. No new failure modes are being
introduced by the proposed change.
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
the margin of safety?
Response: No.
Removal of TS 3/4.4.10 from the TSs does not reduce the controls
that are required to maintain the structural integrity of ASME
[American Society of Mechanical Engineers] Code Class 1, 2, or 3
components. No safety margins are impacted due to the proposed
change.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket Nos. 50-220 and
50-410, Nine Mile Point Nuclear Station Unit Nos. 1 and 2 (NMP 1 and
2), Oswego County, New York
Date of amendment request: February 11, 2009.
Description of amendment request: The proposed amendment would
delete those portions of the Technical Specifications (TSs) superseded
by 10 CFR Part 26, Subpart I. The proposed change is consistent with
Nuclear Regulatory Commission (NRC)-approved Revision 0 to TS Task
Force (TSTF) Change Traveler, TSTF-511-A, ``Eliminate Working Hour
Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.''
The availability of the TS improvement was announced in the Federal
Register (FR) on December 30, 2008 (73 FR 79923) as part of the
consolidated line item improvement process. The licensee concluded that
the no significant hazards consideration determination as presented in
the FR notice is applicable to NMP 1 and 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:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Removal of the
Technical Specification requirements will be performed concurrently
with the implementation of the 10 CFR Part 26, Subpart I,
requirements. The proposed change does not impact the physical
configuration or function of plant structures, systems, or
components (SSCs) or the manner in which the SSCs are operated,
maintained, modified, tested, and inspected. Worker fatigue is not
an initiator of any accident previously evaluated. Worker fatigue is
not an assumption in the consequence mitigation of any accident
previously evaluated. Therefore, it is concluded that this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
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 removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Working hours will
continue to be controlled in accordance with NRC requirements. The
new rule allows for deviations from controls to mitigate or prevent
a condition adverse to safety or as necessary to maintain the
security of the facility. This ensures that the new rule will not
unnecessarily restrict working hours and thereby create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed change does not alter the plant
configuration, require new plant equipment to be installed, alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Working hours will
continue to be controlled in accordance with NRC requirements. The
proposed change does not involve any physical changes to the plants
or alter the manner in which plant systems are operated, maintained,
modified, tested, and inspected. The proposed change does not alter
the manner in which safety limits, limiting safety
[[Page 18256]]
system settings or limiting conditions for operation are determined.
The safety analysis acceptance criteria are not affected by this
change. The proposed change will not result in plant operation in a
configuration outside the design basis. The proposed change does not
adversely affect systems that respond to safely shutdown the plants
and to maintain the plants in a safe shutdown condition. Removal of
plant-specific Technical Specification administrative requirements
will not reduce a margin of safety because the requirements in 10
CFR Part 26 are adequate to ensure that worker fatigue is managed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: February 20, 2009.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) requirements related to control
room envelope habitability in TS 3.7.3, ``Plant Systems Control Room
Emergency Outside Air Supply (CREOAS) System,'' and TS Section 5.5,
``Administrative Controls Programs and Manuals.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on October 17, 2006 (71 FR 61075), on possible
amendments to revise the plant specific TS, to strengthen TS
requirements regarding control room envelope (CRE) habitability by
changing the action and surveillance requirements associated with the
limiting condition for operation operability requirements for the CRE
emergency ventilation system. A new TS administrative controls program
on CRE habitability is being added, including a model safety evaluation
and model no significant hazards consideration determination, using the
consolidated line-item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on January 17,
2007 (72 FR 2022). The licensee affirmed the applicability of the model
NSHC determination in its application dated February 20, 2009.
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 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 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 proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief : Mark Kowal.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: March 23, 2009.
Description of amendment request: The proposed amendment would
delete those portions of the Technical Specifications (TSs) superseded
by Part 26, Subpart I of Title 10 of the Code of Federal Regulations
(10 CFR). This change incorporates NRC approved Revision 0 of Technical
Specification Task Force (TSTF) Improved Standard Technical
Specification Change Traveler, TSTF-511, ``Eliminate Working Hour
Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.''
The availability of this TS improvement was announced as part of the
consolidated line item improvement process (CLIIP) in the Federal
Register on December 30, 2008 (73 FR 79923).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change removes technical specification restrictions
on working hours for personnel who perform safety related functions.
The technical specification
[[Page 18257]]
restrictions are superseded by the worker fatigue requirements in 10
CFR Part 26. Removal of the technical specification requirements
will be performed concurrently with the implementation of 10 CFR
Part 26, Subpart I, requirements. The proposed change does not
impact the physical configuration or function of plant structures,
systems, or components (SSCs) or the manner in which SSCs are
operated, maintained, modified, tested, or inspected. Worker fatigue
is not an assumption in the consequence mitigation of any accident
previously evaluated.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
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 removes technical specification restrictions
on working hours for personnel who perform safety related functions.
The technical specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Working hours will
continue to be controlled in accordance with NRC requirements. The
new rule allows for deviations from controls to mitigate or prevent
a condition adverse to safety or as necessary to maintain the
security of the facility. This ensures that the new rule will not
unnecessarily restrict working hours and thereby create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not alter the plant configuration,
require new plant equipment to be installed, alter accident analysis
assumptions, add any initiators, or effect the function of plant
systems or the manner in which systems are operated, maintained,
modified, tested, or inspected.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change removes technical specification restrictions
on working hours for personnel who perform safety related functions.
The technical specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26.
The proposed change does not involve any physical changes to the
plant or alter the manner in which plant systems are operated,
maintained, modified, tested, or inspected. The proposed change does
not alter the manner in which safety limits, limiting safety system
settings or limiting conditions or operation are determined. The
safety analysis acceptance criteria are not affected by this change.
The proposed change will not result in plant operation in a
configuration outside the design basis. The proposed change does not
adversely affect systems that respond to safely shutdown the plant
and to maintain the plant in a safe shutdown condition.
Removal of plant-specific technical specification administrative
requirements will not reduce a margin of safety because the
requirements in 10 CFR Part 26 are adequate to ensure that worker
fatigue is managed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor,
Baltimore, MD 21202.
NRC Branch Chief: Mark G. Kowal.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: December 29, 2008.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.8.4, ``DC [Direct Current] Sources--
Operating,'' and TS 3.8.5, ``DC Sources--Shutdown.'' Specifically, this
amendment would revise the battery connection resistance limits in
Surveillance Requirement (SR) 3.8.4.2 and SR 3.8.4.5 from 150 micro-
ohms (150E-6 ohm) to 69 micro-ohms (69E-6 ohm). TS 3.8.5 is affected by
virtue of SR 3.8.5.1 invoking both SR 3.8.4.2 and SR 3.8.4.5 for DC
sources that are required to be operable in Modes 5 and 6.
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 replaces a battery surveillance limit with a
value based on voltage drop calculations for each of the four
battery subsystems at Callaway under both normal operating and
accident load profiles. The new value is more conservative, as well
as being more appropriate, as an acceptance criterion for verifying
battery operability pursuant to SR 3.8.4.2 and SR 3.8.4.5, thus
providing greater assurance that the batteries can perform their
specified safety functions with regard to accident mitigation.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no design changes. All design, material, and construction standards
that were applicable prior to this amendment request will be
maintained. There will be no changes to any design or operating
limits.
The proposed change will not adversely affect accident
initiators or precursors, nor adversely alter the design
assumptions, conditions, and configuration of the facility or the
manner in which the plant is operated and maintained. The proposed
change will not alter or prevent the ability of structures, systems,
and components (SSCs) from performing their intended functions to
mitigate the consequences of an initiating event within the assumed
acceptance limits.
The proposed change does not physically alter safety-related
systems nor affect the way in which safety-related systems perform
their functions.
All accident analysis acceptance criteria will continue to be
met with the proposed change. The proposed change will not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of any
accident previously evaluated. The applicable radiological dose
criteria will continue to be 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.
There are no proposed design changes nor are there any changes
in the method by which any safety-related plant structure, system,
or component (SSC) performs its specified safety function. The
proposed changes will not affect the normal method of plant
operation or change any operating parameters. Equipment performance
necessary to fulfill safety analysis missions will be unaffected.
The proposed change will not alter any assumptions required to meet
the safety analysis acceptance criteria.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this amendment. There will be no adverse effect or
challenges imposed on any safety-related system as a result of this
amendment.
The proposed amendment will not alter the design or performance
of the 7300 Process Protection System, Nuclear Instrumentation
System, or Solid State Protection System used in the plant
protection systems.
The proposed change does not, therefore, create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change does not involve a significant
reduction in a margin of safety?
Response: No.
There will be no effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
impact on the overpower limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor (FQ), nuclear
enthalpy rise hot channel factor (F[Delta]H), loss of coolant
accident peak cladding temperature (LOCA PCT), peak local power
density, or any other margin of safety. The applicable radiological
dose consequence acceptance criteria will continue to be met.
[[Page 18258]]
The proposed change does not eliminate any surveillances or
alter the frequency of surveillances required by the Technical
Specifications; however, the acceptance criterion for the specified
battery resistance surveillances will be more restrictive. None of
the acceptance criteria for any accident analysis will be changed.
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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 4, 2009.
Description of amendment request: The proposed amendment consists
of changes to the approved fire protection program as described in Wolf
Creek Generating Station (WCGS) Updated Safety Analysis Report (USAR).
Specifically, a deviation from certain technical requirements of Title
10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix R,
Section III.G.2, as documented in Appendix 9.5E of the WCGS USAR, is
requested regarding the use of operator manual actions in lieu of
meeting circuit separation protection criteria. Table 3-1 of the
submittal dated March 4, 2009 (Agencywide Documents Access and
Management System (ADAMS) Accession No. ML090771269), identifies the
proposed feasible and reliable operator manual actions requested for
permanent approval and Table 3-2 of the submittal identifies the
proposed feasible operator manual actions requested for approval on an
interim basis. The interim operator actions will be eliminated with the
implementation of associated design change package.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design function of structures, systems and components are
not impacted by the proposed change. The proposed change involves
the performance of operator manual actions to achieve and maintain
safe shutdown in the event of a fire outside of the control room and
will not initiate an event. The proposed change does not increase
the probability of occurrence of a fire or any other accident
previously evaluated.
The proposed operator manual actions are feasible and reliable
and demonstrate that the plant can be safely shutdown in the event
of a fire. No significant consequences result from the performance
of the proposed change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The design function of structures, systems and components are
not impacted by the proposed change. The proposed change involves
the performance of operator manual actions to achieve and maintain
safe shutdown in response to a fire outside of the control room. The
operator manual actions do 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 accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
For the permanent operator manual actions, 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 have been verified that they can be performed through
demonstration and the actions are proceduralized. The proposed
actions are feasible and reliable and demonstrate that the plant can
be safely shutdown in the event of a fire.
For the interim operator manual actions adequate time is
available to feasibly perform the proposed operator manual actions
and a compensatory measure fire watch is provided for the affected
area as an added defense in depth fire protection measure.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 6, 2009.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) 5.2.2.d regarding the requirement
to develop and implement administrative procedures to limit the working
hours of personnel who perform safety-related functions. The
requirements of TS 5.2.2 have been superseded by Title 10 of the Code
of Federal Regulations (10 CFR) Part 26, Subpart I. The change is
consistent with U.S. Nuclear Regulatory Commission (NRC)-approved
Revision 0 to Technical Specification Task Force (TSTF) Improved
Technical Specification Change Traveler, TSTF-511, ``Eliminate Working
Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part
26.''
The NRC staff issued a ``Notice of Availability of Model Safety
Evaluation, Model No Significant Hazards Determination, and Model
Application for Licensees That Wish To Adopt TSTF-511, Revision 0,
`Eliminate Working Hour Restrictions From TS 5.2.2 To Support
Compliance With 10 CFR Part 26,' '' in the Federal Register on December
30, 2008 (73 FR 79923). The notice included a model safety evaluation,
a model no significant hazards consideration (NSHC) determination, and
a model license amendment request, using the consolidated line item
improvement process. In its application dated March 6, 2009, 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 determination 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 removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Removal of the
Technical Specification requirements will be performed concurrently
with the implementation of the 10 CFR Part
[[Page 18259]]
26, Subpart I, requirements. The proposed change does not impact the
physical configuration or function of plant structures, systems, or
components (SSCs) or the manner in which SSCs are operated,
maintained, modified, tested, or inspected. Worker fatigue is not an
initiator of any accident previously evaluated. Worker fatigue is
not an assumption in the consequence mitigation of any accident
previously evaluated. Therefore, it is concluded that this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
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 removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Working hours will
continue to be controlled in accordance with NRC requirements. The
new rule allows for deviations from controls to mitigate or prevent
a condition adverse to safety or as necessary to maintain the
security of the facility. This ensures that the new rule will not
unnecessarily restrict working hours and thereby create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed change does not alter the plant
configuration, require new plant equipment to be installed, alter
accident analysis assumptions, add any initiators, or [a]ffect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. The proposed change
does not involve any physical changes to plant or alter the manner
in which plant systems are operated, maintained, modified, tested,
or inspected. The proposed change does not alter the manner in which
safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
change will not result in plant operation in a configuration outside
the design basis. The proposed change does not adversely affect
systems that respond to safely shutdown the plant and to maintain
the plant in a safe shutdown condition. Removal of plant-specific
Technical Specification administrative requirements will not reduce
a margin of safety because the requirements in 10 CFR Part 26 are
adequate to ensure that worker fatigue is managed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management 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].
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: June 23, 2008.
Brief description of amendments: This request modifies the subject
Technical Specifications (TSs) and Bases by changing the logic
configuration of TS Table 3.3.2-1, ``Engineered Safety Feature
Actuation System Instrumentation,'' Function 5.b.(5), ``Turbine Trip
and Feedwater Isolation, Feedwater Isolation, Doghouse Water Level--
High High.'' The existing one-out-of-one (1/1) logic per train per
doghouse is being modified to a two-out-of-three (2/3) logic per train
per doghouse. The proposed change will improve the overall reliability
of this function and will reduce the potential for spurious actuations.
Date of issuance: April 2, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 249/243.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: February 24, 2009 (74
FR 8276).
The Commission's related evaluation, state consultation, and final
no significant hazards consideration determination of the amendments
are contained in a Safety Evaluation dated April 2, 2009.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: March 20, 2008.
Brief description of amendments: The proposed amendments would
revise the McGuire licensing basis by adopting the Alternative Source
Term (AST) radiological analysis methodology as
[[Page 18260]]
allowed by 10 CFR 50.67, ``Accident source term,'' for the Loss of
Coolant Accident. This amendment request represents full scope
implementation of the AST as described in Nuclear Regulatory Commission
(NRC) Regulatory Guide 1.183, ``Alternative Radiological Source Terms
for Evaluating Design Basis Accidents at Nuclear Power Reactors,
Revision 0.'' Selective implementation of AST for the McGuire Fuel
Handling Accidents was approved by the NRC on December 22, 2006. There
are no changes proposed to the McGuire Technical Specifications within
this amendment request. The application of the AST methodology to the
Loss of Coolant Accident (LOCA) radiological analysis will allow
McGuire to resolve the Control Room envelope degraded boundary
condition as discussed in McGuire's response to NRC Generic Letter
2003-01, ``Control Room Habitability,'' dated February 19, 2004.
By separate amendment request dated January 22, 2008, Duke proposed
to revise the McGuire Technical Specification (TS) requirements related
to control room envelope habitability in TS 3.7.9, ``Control Room Area
Ventilation System.'' The proposed changes are consistent with the
Industry and NRC-approved Technical Specification Task Force (TSTF)
change TSTF-448, Control Room Habitability, Revision 3 and the NRC
Consolidated Line Item Improvement Process (CLIIP).
Duke has performed a review of all McGuire License Amendment
Requests (LAR) currently under review by the NRC for impacts to this
AST LAR. None of these LARs impact any assumptions or results of the
LOCA AST radiological analysis.
Date of issuance: March 31, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 251 and 231.
Renewed Facility Operating License Nos. NPF-9 and NPF-17: The
amendments revised the license.
Public comments requested as to proposed no significant hazards
consideration (NSHC): The notice provided an opportunity to submit
comments on the Commission's proposed NSHC determination by March 30,
2009. No comments have been received to date. However, the notice also
provided an opportunity to request a hearing by April 28, 2009, but
indicated that if the Commission make a final NSHC determination, any
such hearing would take place after issuance of the amendment.
Date of initial notice in Federal Register: February 27, 2009 (74
FR 9009).
The supplements dated May 28, 2008, October 6, 2008, December 17,
2008 and February 12, 2009, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2009.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: December 8, 2008.
Brief description of amendment: The amendment added a license
condition to allow a one-time extension of surveillance requirements
involving the 18-month channel calibration and logic system functional
tests for one channel of the reactor water level instrumentation
system. The extension is to account for the effects of rescheduling the
next refueling outage from early to late 2009.
Date of issuance: April 1, 2009.
Effective date: As of the date of issuance and shall be implemented
within 15 days from the date of issuance.
Amendment No.: 162.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 27, 2008 (74 FR
4770).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 2009.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: July 21, 2008.
Brief description of amendment: The amendment deleted the exception
to Limiting Condition for Operation (LCO) 3.0.4 to the 30-day allowable
outage time of the Startup No. 2 Transformer and corrected a spelling
error in Technical Specification (TS) 3.8.1. The NRC approved the
adoption of Industry/TS Task Force (TSTF) change TSTF-359, ``Increased
Flexibility in Mode Restraints,'' for ANO-1 in TS Amendment 232 dated
April 2, 2008 (Agencywide Documents Access and Management System
(ADAMS) Accession No. ML080600006). The intent of TSTF-359 was to
eliminate exceptions to LCO 3.0.4 within individual specifications and
provide requirements within LCO 3.0.4 to control mode changes when TS-
required equipment is inoperable. The licensee omitted deleting this
LCO 3.0.4 exception in its October 22, 2007 (ADAMS Accession No.
ML073030542), amendment request to adopt TSTF-359.
Date of issuance: March 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 236.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications/license.
Date of initial notice in Federal Register: October 21, 2008 (73 FR
62563).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 30, 2009.
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: December 16, 2008, as
supplemented by letter dated February 19, 2009.
Brief description of amendment: This amendment request would revise
the Technical Specifications Section 2.1.2, Safety Limit Minimum
Critical Power Ratio (SLMCPR) for two-loop and single-loop operation.
Date of issuance: March 26, 2009.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 232.
Facility Operating License No. DPR-35: The amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: January 23, 2009 (74 FR
4250).
The supplemental letter dated February 19, 2009, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated March 26, 2009.
No significant hazards consideration comments received: No.
[[Page 18261]]
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois; Docket Nos.
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois
Date of application for amendment: April 9, 2008, as supplemented
by letter dated October 1, 2008.
Brief description of amendment: The amendments revise Technical
Specifications (TSs) 5.5.6, Pre-Stressed Concrete Containment Tendon
Surveillance Program, and 5.6.8, Tendon Surveillance Report, for
consistency with the requirements of Title 10 Code of Federal
Regulations (10 CFR) Section 50.55a, Codes and standards, paragraph
(g)(4) for components classified as American Society of Mechanical
Engineers Boiler and Pressure Vessel Code (ASME Code) Class CC, by
replacing the reference to the specific ASME Code year for the tendon
surveillance program with a requirement to use the applicable ASME Code
and addenda as required by 10 CFR 50.55a.
Date of issuance: March 26, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: Braidwood Unit 1-158; Braidwood Unit 2-158; Byron
Unit No. 1-163; and Byron Unit No. 2-163.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66:
The amendments revise the TSs and Licenses.
Date of initial notice in Federal Register: July 1, 2008 (73 FR
37504).
The October 1, 2008, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 2009.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station (CPS), Unit No. 1, DeWitt County, Illinois
Date of application for amendment: September 2, 2008.
Brief description of amendment: The amendment requested to amend
the CPS Unit No. 1 Technical Specifications (TS) to relocate the TS
surveillance requirement (SR) 3.8.3.6 from the TS to a licensee-
controlled document. SR 3.8.3.6 requires the emergency diesel generator
fuel oil storage tanks to be drained, sediment removed, and cleaned on
a 10-year interval. The request is submitted consistent with the
guidance contained in Nuclear Regulatory Commission (NRC)-approved
Technical Specifications Task Force Report 2 (TSTF-2).
Date of issuance: April 2, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 186.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
65687) and January 27, 2009 (74 FR 4771). The notice on January 27,
2009, was inadvertently placed in the Federal Register a second time
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 2, 2009.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit No. 2 (BVPS-2), Beaver County,
Pennsylvania
Date of application for amendment: November 7, 2008.
Brief description of amendment: The amendment modifies the method
used to calculate the available net positive suction head (NPSH) for
the BVPS-2 recirculation spray (RS) pumps as described in the BVPS-2
Updated Final Safety Analysis Report (UFSAR). The BVPS-2 UFSAR takes
credit for containment overpressure by allowing for the difference
between containment total pressure and the vapor pressure of the water
in the containment sump in the available NPSH calculation.
Date of issuance: March 26, 2009.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 167.
Facility Operating License No. NPF-73: The amendment revised the
License and the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: December 16, 2008 (73
FR 76411).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2009.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2 (CNP-1 and CNP-2), Berrien County,
Michigan
Date of application for amendment: October 21, 2008.
Brief description of amendment: The amendment modifies Technical
Specification 5.6.3, ``Radioactive Effluent Release Report,'' by
changing the required annual submittal date for the report from
``within 90 days of January 1 of each year'' (i.e., prior to April 1),
to ``prior to May 1 of each year.''
Date of issuance: March 30, 2009.
Effective date: As of the date of issuance.
Amendment Nos.: 308 (CNP-1), 290 (CNP-2).
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Renewed Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: December 16, 2008 (73
FR 76412).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 30, 2009.
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: April 22, 2008, as supplemented by
letter dated March 6, 2009.
Brief description of amendment: The amendment modifies the
Technical Specification (TS) 2.7, ``Electrical Systems,'' Limiting
Condition for Operation (LCO) 2.7(2)j related to the allowed outage
time for the Emergency Diesel Generators (EDGs). The change clarifies
LCO 2.7(2)j such that a single period of inoperability for one EDG is
limited to 7 consecutive days and that the cumulative total time of
inoperability for both EDGs during any calendar month cannot exceed 7
days.
Date of issuance: March 27, 2009.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 258.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: June 17, 2008 (73 FR
34342). The supplemental letter dated March 6, 2009, provided
additional information that clarified the application, did not expand
the scope of the application as
[[Page 18262]]
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 March 27, 2009.
No significant hazards consideration comments received: No.
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 application for amendments: April 3, 2008, as supplemented
by letters dated June 20, October 1, November 6, and December 16, 2008.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to
remove Surveillance Requirement (SR) 3.7.5.6, and revised TS 3.7.6,
``Condensate Storage Tank (CST) and Fire Water Storage Tank (FWST),''
to remove the FWST level requirements, revise the CST level
requirements, and revise TS 3.7.6 to be consistent with the NUREG-1431,
``Standard Technical Specifications (STS).'' Specifically, these
changes reflect design changes made to the CSTs and are necessary to
support the on-line refurbishment of the FWST and replacement of the
recirculation piping for the fire water pumps. The design changes to
the CSTs are intended to eliminate the reliance on the FWST for
additional seismically-qualified feedwater supply and thus, make the
existing TS requirements for the FWST unnecessary.
Date of issuance: March 30, 2009.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1-204; Unit 2-205.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: July 29, 2008 (78 FR
43956). The supplemental letters dated June 20, October 1, November 6,
and December 16, 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 March 30, 2009.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of application for amendment: September 18, 2008.
Brief description of amendment: The amendment revised WBN Unit 1
Technical Specification 3.8.7, ``Inverters--Operating.'' The amendment
revised the requirement to two inverters for each of the four channels.
Date of issuance: March 24, 2009.
Effective date: As of the date of issuance and shall be implemented
within 240 days of issuance.
Amendment No.: 76.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specification 3.8.7 and Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
65697).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 24, 2009.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 14, 2007, as supplemented by
letters dated April 18, May 9, June 15, August 31, September 12 and 20,
October 16, November 16, two letters dated December 14, and December
18, 2007; two letters dated January 18, January 31, February 26 and 28,
March 14, April 26, May 14, June 19, and July 31, 2008; and January 16
and 29, and February 17 and 27, 2009.
Brief description of amendment: The amendment revised the licensing
basis for the Main Steam and Feedwater Isolation System (MSFIS)
controls to incorporate field programmable gate array technology. Other
related changes requested in the March 14, 2007, application were
previously approved in Amendment No. 174, dated August 28, 2007,
Amendment No. 175, dated March 3, 2008, Amendment No. 176, dated March
21, 2008, and Amendment No. 177, dated April 3, 2008.
Date of issuance: March 31, 2009.
Effective date: Effective as of date of issuance and shall be
implemented before entry into Mode 3 in the restart from Refueling
Outage 17.
Amendment No.: 181.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License.
Date of initial notice in Federal Register: June 19, 2007 (72 FR
33785). The supplemental letters dated April 18, May 9, June 15, August
31, September 12 and 20, October 16, November 16, two letters dated
December 14, and December 18, 2007; two letters dated January 18,
January 31, February 26 and 28, March 14, April 26, May 14, June 19,
and July 31, 2008; and January 16 and 29, and February 17 and 27, 2009,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 2009.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 10th day of April 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-8832 Filed 4-20-09; 8:45 am]
BILLING CODE 7590-01-P