[Federal Register Volume 74, Number 95 (Tuesday, May 19, 2009)]
[Notices]
[Pages 23440-23452]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-11268]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0204]
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 April 23, 2009, to May 6, 2009. The last
biweekly notice was published on May 5, 2009 (74 FR 20741).
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
[[Page 23441]]
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.
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 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:
[[Page 23442]]
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) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: February 19, 2009.
Description of amendment request: The amendments would relocate the
reactor coolant system pressure and temperature (P/T) limits and the
low temperature overpressure protection (LTOP) enable temperatures to a
licensee-controlled document outside of the Technical Specifications
(TSs). The P/T limits and LTOP enable temperatures would be specified
in a Pressure and Temperature Limits Report (PTLR) that would be
located in the Palo Verde Nuclear Generating Station (PVNGS) Technical
Requirements Manual and administratively controlled by a new TS 5.6.9.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This proposed change revises the Technical Specifications by
relocating the reactor coolant system (RCS) pressure and temperature
limits, heatup and cooldown curves and low temperature overpressure
protection (LTOP) enable temperatures from the Technical
Specifications to an [Arizona Public Service] APS-controlled RCS
Pressure and Temperature Limits Report (PTLR), and requiring that
the limits in the PTLR be determined using the analytical methods
described in the NRC-approved Topical Report CE NPSD-683-A.
Relocation of this information and updating it using NRC-approved
methodology will not alter the requirement to update the RCS
pressure and temperature curves and limits in accordance with 10 CFR
50 Appendices G and H. Updating the P/T curves and LTOP limits
ensures the reactor coolant system's pressure boundary integrity is
protected throughout plant life. Consequently, this proposed change
is determined to not contribute to an increase in the probability
of, or the initiation of, a design basis accident. Similarly, the
safety analysis information presented in the Updated Final Safety
Analysis Report remains unchanged.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change revises the Technical Specifications by
relocating the RCS pressure and temperature limits, heatup and
cooldown curves and LTOP enable temperatures from the Technical
Specifications to a PVNGS PTLR, and requiring that the limits in the
PTLR be determined using the analytical methods described in the
NRC-approved Topical Report CE NPSD-683-A. The PTLR documents
removal, testing and analyzing the surveillance capsules, and will
be updated by APS to reflect the results of testing and analysis of
surveillance specimens withdrawn in the future. Removal, testing and
analysis of surveillance specimens may result in a need to implement
changes to the RCS pressure and temperature limits. Such changes are
implemented to ensure the integrity of the RCS pressure boundary
throughout plant lifetime. Updates to the RCS pressure and
temperature curves and limits will not create a new or different
kind of accident. Relocating the P/T curves, heatup and cooldown
rates and LTOP limits to the PTLR has no impact on any safety
analyses.
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.
Pressure and temperature curves and limits are provided as
limits to plant operation to ensure RCS pressure boundary integrity
is maintained throughout the plant's lifetime. Changes to the RCS
pressure and temperature curves and limits, resulting from the
removal, testing and analysis of surveillance capsules, are only
made within the acceptable margin limits thereby maintaining the
required margin of safety. There is no change to the safety
analysis.
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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
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: December 1, 2008.
Description of amendment request: The proposed amendments would
correct a non-conservative Technical Specification (TS) Surveillance
Requirement by revising McGuire TS 3.8.1.4 to increase the minimum
required amount of fuel oil for the Emergency Diesel Generators fuel
oil day tank as read on the local fuel gauge used to perform the
surveillance.
[[Page 23443]]
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.
Implementation of the proposed amendment does not significantly
increase the probability or the consequences of an accident
previously evaluated. The Emergency Diesel Generators (EDGs) and
their associated emergency buses function as accident mitigators.
The proposed changes do not involve a change in the operational
limits or the design of the electrical power systems (particularly
the emergency power systems) or change the function or operation of
plant equipment or affect the response of that equipment when called
upon to operate. The proposed change to TS SR 3.8.1.4 confirms the
minimum supply of fuel oil in the emergency diesel generators (EDG)
fuel oil day tank. The minimum value for the affected parameter is
being increased in the conservative direction and further ensures
the EDGs ability to fulfill their safety related function. Thus,
based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve a change in the operational
limits or the design capabilities of the emergency electrical power
systems. The proposed changes do not change the function or
operation of plant equipment or introduce any new failure
mechanisms. The evaluation that supports this LAR included a review
of the EDG fuel oil system to which this parameter applies. The
proposed changes do not introduce any new or different types of
failure mechanisms; plant equipment will continue to respond as
designed and analyzed.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
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 the fuel cladding, the reactor coolant
system and the containment system will not be adversely impacted by
the proposed changes. Thus, it is concluded that the proposed TS and
TS Basis changes do 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: 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 Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: March 5, 2009.
Description of amendment request: The proposed amendment will
revise the Reactor Vessel Heatup, Cooldown, and Low Temperature
Overpressure Protection curves in Technical Specifications (TSs) 3.4.3
and 3.4.12 to incorporate the most recent estimates of lifetime neutron
fluence and the effects of the Stretch Power Uprate (Amendment No.
241).
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. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated.
The proposed TS changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
There are no physical changes to the plant being introduced by the
proposed changes to the heatup and cooldown limitation curves. The
proposed changes do not modify the RCS [Reactor Coolant System]
pressure boundary. That is, there are no changes in operating
pressure, materials, or seismic loading. The proposed changes do not
adversely affect the integrity of the RCS pressure boundary such
that its function in the control of radiological consequences is
affected. The proposed heatup and cooldown limitation curves were
generated in accordance with the fracture toughness requirements of
10 CFR 50 [Title 10 of the Code of Federal Regulations Part 50]
Appendix G, and ASME B&PV code [American Society of Mechanical
Engineers Boiler and Pressure Vessel Code], Section XI, Appendix G
edition with 2000 Addenda. The proposed heatup and cooldown
limitation curves were established in compliance with the
methodology used to calculate and predict effects of radiation on
embrittlement of RPV [Reactor Pressure Vessel] beltline materials.
Use of this methodology provides compliance with the intent of 10
CFR 50 Appendix G and provides margins of safety that ensure non-
ductile failure of the RPV will not occur. The proposed heatup and
cooldown limitation curves prohibit operation in regions where it is
possible for non-ductile failure of carbon and low alloy RCS
materials to occur. Hence, the primary coolant pressure boundary
integrity will be maintained throughout the limit of applicability
of the curves, 29.2 EFPY [Effective Full-Power Years].
Operation within the proposed LTOPS [Low Temperature
Overpressure Protection System] limits ensures that
overpressurization of the RCS at low temperatures will not result in
component stresses in excess of those allowed by the ASME B&PV Code
Section XI Appendix G.
Consequently, the proposed changes do not involve a significant
increase in the probability or the consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No new modes of operation are introduced by the proposed
changes. The proposed changes will not create any failure mode not
bounded by previously evaluated accidents. Further, the proposed
changes to the heatup and cooldown limitation curves and the LTOPS
limits do not affect any activities or equipment other than the RCS
pressure boundary and do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Consequently, the proposed changes do not involve a significant
increase in the probability or consequence of a new or different
kind of accident, from any accident previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
The proposed TS changes do not involve a significant reduction
in the margin of safety.
The revised heatup and cooldown limitation curves and LTOPS
limits are established in accordance with current regulations and
the ASME B&PV Code 1998 edition with 2000 Addenda. These proposed
changes are acceptable because the ASME B&PV Code maintains the
margin of safety required by 10 CFR 50.55(a). Because operation will
be within these limits, the RCS materials will continue to behave in
a non-brittle manner consistent with the original design bases.
The proposed changes to the allowable operation of charging and
safety injection pumps when LTOPS is required to be operable is
consistent with the IP2 licensing bases as established in TS
Amendment 224.
Therefore, Entergy has concluded that the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 23444]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: Richard V. Guzman.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: March 25, 2009.
Description of amendment request: The proposed amendment would add
two Emergency Core Cooling System (ECCS) valves to Surveillance
Requirement (SR) 3.5.2.1. The SR is designed to verify that ECCS valves
whose single failure could cause loss of the ECCS function are in the
required position with ac power removed so that misalignment or single
failure cannot prevent completion of the ECCS function. Entergy plans
to install an alternate source of power during the spring 2010
refueling outage to provide the required position indication. The
proposed changes support Entergy's resolution to Generic Letter (GL)
2004-02 by establishing a licensing basis that supports meeting the
regulatory requirements of the GL.
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 adds two ECCS valves to SR 3.5.2.1. The
purpose of the surveillance is to assure that the valves are in
their required position with normal ac power removed from the valve
operator so that misalignment or single failure cannot prevent
completion of the ECCS function. The performance of the SR does not
involve any actions related to the initiation of an accident and
therefore the proposed changes cannot increase the probability of an
accident. Misalignment or single failure of one of the two valves
being added to TS [Technical Specifications] could cause a loss of
the ECCS function based on GSI [Generic Safety Issue]-191
evaluations, so the change will not increase the consequences of an
accident but rather provide assurance that no such increase can
occur. Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response--No.
The proposed change adds two ECCS valves to SR 3.5.2.1. The
purpose of the surveillance is to assure that the valves are in
their required position with normal ac power removed from the valve
operators so that misalignment or single failure cannot prevent
completion of the ECCS function. The provision of alternate power to
the existing valve position indication during the upcoming spring
2010 outage (2R19), will allow the valve operators to be normally
deenergized. The change assures that the valves will be in their
correct position and does not introduce any new failure modes or the
possibility of a different 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 change involve a significant reduction in a
margin of safety?
Response--No.
The proposed change adds two ECCS valves to SR 3.5.2.1. The
purpose of the surveillance is to assure that the valves are in
their required position with normal ac power removed so that
misalignment or single failure cannot prevent completion of the ECCS
function. The valves will be re-energized 24 hours following a DBA
[design-basis accident] and therefore will be capable of performing
their required function of isolating a potential passive failure at
that time. This ensures that the ECCS function can be performed
without a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: Richard V. Guzman.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: March 29, 2009.
Description of amendment request: The proposed amendment will
establish a more restrictive acceptance criterion for surveillance
requirement (SR) 3.8.6.6 regarding periodic verification of capacity
for the affected station batteries.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change revises the acceptance criterion applied
to an existing surveillance test for the Indian Point 2 station
batteries. Performing a technical specification surveillance test is
not an accident initiator and does not increase the probability of
an accident occurring. The proposed revision to the test acceptance
criterion is based on the design calculation for battery performance
at the minimum design temperature. The proposed new value for the
test acceptance criteria is more limiting than the existing value
which does not account for the minimum environmental design
temperature assumed for the limiting battery locations. Establishing
a test acceptance criterion that bounds existing or assumed
conditions validates the equipment performance assumptions used in
the accident mitigation safety analyses. 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?
No. The proposed change revises the test acceptance criterion
for an existing technical specification surveillance test conducted
on the existing station batteries. The proposed change does not
involve installation of new equipment or modification of existing
equipment, so that no new equipment failure modes are introduced.
Also, the proposed change in test acceptance criterion does not
result in a change to the way that the equipment or facility is
operated so that no new accident initiators are created. Therefore
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The conduct of performance tests on safety-related plant
equipment is a means of assuring that the equipment is capable of
performing its intended safety function and therefore maintaining
the margin of safety established in the safety analysis for the
facility. The proposed change in the acceptance criterion for the
battery capacity surveillance test is more conservative and more
restrictive than the value currently in the technical specification
and is based on the applicable design calculation for these
components.
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
[[Page 23445]]
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: Richard V. Guzman.
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 amendment request: March 26, 2009.
Description of amendment request: The proposed amendments would
revise the fire protection program (FPP) to eliminate the requirement
for the backup manual carbon dioxide (CO2) fire suppression
system in the upper cable spreading rooms.
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 the FPP to eliminate the requirement
to maintain the backup CO2 fire suppression system for
the upper cable spreading rooms. With the exception of the
CO2 fire suppression system itself, the proposed change
does not result in any physical changes to safety related
structures, systems, or components [SSCs], or the manner in which
they are operated, maintained, modified, tested, or inspected. The
proposed change does not degrade the performance or increase the
challenges of any safety related SSCs assumed to function in the
accident analysis. The proposed change does not change the
probability of a fire occurring since the fire ignition frequency is
independent of the method of fire suppression. The proposed change
does not affect the consequences of an accident previously evaluated
since the fire safe shutdown analysis assumes fire damage throughout
the affected fire area. The results of a fire in the upper cable
spreading room would only affect one engineered safety features
division. Sufficient redundancy exists in the engineered safety
features fed from the other division to achieve a reactor shutdown
and to maintain the reactor in a safe shutdown condition.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the FPP to eliminate the requirement
to maintain the backup CO2 fire suppression system for
the upper cable spreading rooms. With the exception of the
CO2 fire suppression system itself, the proposed change
does not result in any physical changes to safety related
structures, systems, or components, or the manner in which they are
operated, maintained, modified, tested, or inspected. The proposed
change does not degrade the performance or increase the challenges
of any safety related SSCs assumed to function in the accident
analysis. As a result, the proposed change does not introduce nor
increase the number of failure mechanisms of a new or different type
than those previously evaluated. The fire safe shutdown analysis
assumes fire damage throughout the area consistent with a complete
lack of fire suppression capability. Potential habitability hazards
associated with actuation of the CO2 system are
eliminated with the proposed change.
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 the FPP to eliminate the requirement
to maintain the backup CO2 fire suppression system for
the upper cable spreading rooms. With the exception of the
CO2 fire suppression system itself, the proposed change
does not result in any physical changes to safety related
structures, systems, or components, or the manner in which they are
operated, maintained, modified, tested, or inspected. The proposed
change does not degrade the performance or increase the challenges
of any safety related SSCs assumed to function in the accident
analysis. Since the backup manual CO2 fire suppression
system is not credited in the safe shutdown analysis to protect the
upper cable spreading rooms, the proposed change does not impact
plant safety since the conclusions of the fire safe shutdown
analysis remain unchanged.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County,
Texas
Date of amendment request: April 1, 2009.
Brief description of amendments: The proposed amendment would
delete Technical Specification (TS) 5.2.2.d, in TS 5.2.2, ``Unit
Staff,'' 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.d 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. In its application dated April 1,
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 adopted by the licensee, is presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change removes TS restrictions on working hours for
personnel who perform safety related functions. The TS restrictions
are superseded by the worker fatigue requirements in 10 CFR 26.
Removal of the TS requirements will be performed concurrently with
the implementation of the 10 CFR 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
[[Page 23446]]
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 TS restrictions on working hours for
personnel who perform safety related functions. The TS restrictions
are superseded by the worker fatigue requirements in 10 CFR 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 TS restrictions on working hours for
personnel who perform safety related functions. The TS restrictions
are superseded by the worker fatigue requirements in 10 CFR 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 TS administrative requirements will
not reduce a margin of safety because the requirements in 10 CFR 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: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Michael T. Markley.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County,
Texas
Date of amendment request: April 2, 2009.
Brief description of amendments: The amendment revises Technical
Specification (TS) 3.3.1, ``Reactor Trip System (RTS)
Instrumentation,'' to add Surveillance Requirement (SR) 3.3.1.16 to
Function 3 of TS Table 3.3.1-1. SR 3.3.1.16 requires that RTS RESPONSE
TIMES be verified to be within limits every 18 months on a STAGGERED
TEST BASIS. Function 3 is the power range neutron flux--high positive
rate reactor trip function (hereafter referred to as the positive flux
rate trip (PFRT) function). This change is based on a reanalysis of the
Rod Cluster Control Assembly Bank Withdrawal at Power event.
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 change imposes additional surveillance requirements
to assure safety related structures, systems, and components are
verified to be consistent with the safety analysis and licensing
basis. In this specific case, a response time verification
requirement will be added to the positive flux rate trip (PFRT)
function.
Overall protection system performance will remain within the
bounds of the accident analysis since there are no hardware changes.
The design of the Reactor Trip System (RTS) instrumentation,
specifically the positive flux rate trip (PFRT) function, will be
unaffected. The reactor protection system will continue to function
in a manner consistent with the plant design basis. All design,
material, and construction standards that were applicable prior to
the request are maintained.
The proposed changes will not modify any system interface. The
proposed changes will not affect the probability of any event
initiators. There will be no degradation in the performance of or an
increase in the number of challenges imposed on safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance. The proposed changes will not alter any
assumptions or change any mitigation actions in the radiological
consequences evaluations in the updated Final Safety Analysis Report
(FSAR).
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter 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
changes do not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. The proposed
changes are consistent with safety analysis assumptions and
resultant consequences.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change imposes additional surveillance requirements
to assure safety related structures, systems, and components are
verified to be consistent with the safety analysis and licensing
basis.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. This change will not affect the normal method of plant
operation or change any operating parameters. No performance
requirements will be affected; however, the proposed change does
impose additional surveillance requirements. The additional
requirements are consistent with assumptions made in the safety
analysis and licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change imposes additional surveillance requirements
to assure safety related structures, systems, and components are
verified to be consistent with the safety analysis and licensing
basis.
The proposed changes do not affect the acceptance criteria for
any analyzed event. The margin of safety is affected in that in the
new analyses of the Rod (Bank) Withdrawal at Power analyses, it is
necessary to credit a previously uncredited reactor trip function in
an analysis. However, that reactor trip function is described in the
plant Technical
[[Page 23447]]
Specifications with well-defined operability requirements. An
additional attribute, specifically the channel response time
verification on, a periodic frequency, provides additional assurance
that the trip function performs as credited in the accident
analysis. With the credit for this reactor trip function, all
relevant event acceptance criteria continue to be met. None of the
event acceptance limits are exceeded, and none of the event
acceptance limits are revised by the proposed activity. There is no
effect on the manner in which safety limits, limiting safety system
settings, or limiting conditions for operation are determined nor is
there any effect on those plant systems necessary to assure the
accomplishment of protection functions. There is no impact on the
overpower limit, the minimum departure from nucleate boiling ratio
limit, the radial and axial peaking factor limits, the loss of
coolant accident (LOCA) peak clad temperature limit, nor any other
limit which, in whole or in part, defines a margin of safety. The
radiological dose consequence acceptance criteria listed in the
Standard Review Plan will continue to be met.
Therefore the proposed change does 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: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Michael T. Markley.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP 1), Oswego County, New York
Date of amendment request: March 3, 2009.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) Section 3.2.1, ``Reactor Vessel
Heatup and Cooldown Rates,'' and Section 3.2.2, ``Minimum Reactor
Vessel Temperature for Pressurization,'' by replacing the existing
reactor vessel heatup and cooldown rate limits and the pressure and
temperature limit curves with references to the Pressure and
Temperature Limits Report (PTLR). In addition, a new definition for the
PTLR would be added to TS Section 1.0, ``Definitions,'' and a new
section addressing administrative requirements for the PTLR would be
added to TS Section 6.0, ``Administrative Controls.'' The proposed
changes are consistent with the guidance in Generic Letter 96-03,
``Relocation of the Pressure Temperature Limit Curves and Low
Temperature Overpressure Protection System Limits,'' as supplemented by
TS Task Force (TSTF) traveler TSTF-419-A, ``Revise PTLR Definition and
References in ISTS 5.6.6, RCS [Reactor Coolant System] PTLR.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes modify the TS by replacing references to
existing reactor vessel heatup and cooldown rate limits and P-T
[pressure-temperature] limit curves with references to the PTLR. The
proposed amendment also adopts the NRC-approved methodology of SIR-
05-044-A for the preparation of NMP1 P-T limit curves. In 10 CFR 50
Appendix G, requirements are established to protect the integrity of
the reactor coolant pressure boundary (RCPB) in nuclear power
plants. Implementing the NRC-approved methodology for calculating P-
T limit curves and relocating those curves to the PTLR provide an
equivalent level of assurance that RCPB integrity will be
maintained, as specified in 10 CFR 50 Appendix G.
The proposed changes do not adversely affect accident initiators
or precursors, and do not alter the design assumptions, conditions,
or configuration of the plant or the manner in which the plant is
operated and maintained. The ability of structures, systems, and
components to perform their intended safety function is not altered
or prevented by the proposed changes, and the assumptions used in
determining the radiological consequences of previously evaluated
accidents are not affected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The change in methodology for calculating P-T limits and the
relocation of those limits to the PTLR do not alter or involve any
design basis accident initiators. RCPB integrity will continue to be
maintained in accordance with 10 CFR 50 Appendix G, and the assumed
accident performance of plant structures, systems and components
will not be affected. These changes do not involve any physical
alteration of the plant (i.e., no new or different type of equipment
will be installed), and installed equipment is not being operated in
a new or different manner. Thus, no new failure modes are
introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not affect the function of the RCPB or
its response during plant transients. By calculating the P-T limits
using NRC-approved methodology, adequate margins of safety relating
to RCPB integrity are maintained. The proposed changes do not alter
the manner in which safety limits, limiting safety system settings,
or limiting conditions for operation are determined, there are no
changes to the setpoints at which actions are initiated, and the
operability requirements for equipment assumed to operate for
accident mitigation are not affected.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Acting Branch Chief: John P. Boska.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP 2), Oswego County, New York
Date of amendment request: March 9, 2009.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) testing frequency for the
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.''
Specifically, the proposed change is based on TS Task Force (TSTF)
change traveler TSTF-460-A, Revision 0, and extends the frequency for
testing control rod scram time testing in SR 3.1.4.2 from every 120
days of cumulative Mode 1 operation to 200 days of cumulative Mode 1
operation. A notice of availability of this proposed TS change using
the consolidated line item improvement process was published in the
Federal Register on August 23, 2004 (69 FR 51864). The licensee
affirmed the applicability of the model no significant hazards
consideration determination in its application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 23448]]
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The frequency
of surveillance testing is not an initiator of any accident
previously evaluated. The frequency of surveillance testing does not
affect the ability to mitigate any accident previously evaluated, as
the tested component is still required to be operable.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change does not result in any new or different modes of plant
operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
4. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change continues to test the control rod scram time to ensure the
assumptions in the safety analysis are protected.
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 Acting Branch Chief: John P. Boska.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2 (PINGP), Goodhue County,
Minnesota
Date of amendment request: March 5, 2009, as supplemented by letter
dated April 13, 2009.
Description of amendment request: The proposed amendments would
make changes to the PINGP Technical Specifications (TSs) to revise TS
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SR) 3.8.1.8
Frequency to allow use of the SR 3.0.2 interval extension (1.25 times
the specified 24 month Frequency). This would be an exception to the SR
3.0.2 limitations in the PINGP TS, which do not allow use of the
interval extension for SRs with a 24 month Frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes to add a Frequency Note
to Surveillance Requirement 3.8.1.8 which will allow application of
the Surveillance Requirement 3.0.2 interval extension (1.25 times
the specified 24 month Frequency) for performance of this
surveillance. This would be an exception to the limitations
specified in the Prairie Island Nuclear Generating Plant Technical
Specification Surveillance Requirement 3.0.2 for Surveillance
Requirements with a 24 month Frequency and would allow an interval
up to 30 months for performance of the surveillance.
The emergency diesel generators are not accident initiators and
therefore, these changes do not involve a significant increase [in]
the probability of an accident.
Failure of the bypass relay, by itself, does not prevent an
emergency diesel generator from performing its safety related
functions. Since the accident analyses only require one of the two
trains of onsite emergency AC to be operable, the changes proposed
in the license amendment request do not involve a significant
increase in the consequences of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This license amendment request proposes to add a Frequency Note
to Surveillance Requirement 3.8.1.8 which will allow application of
the Surveillance Requirement 3.0.2 interval extension (1.25 times
the specified 24 month Frequency) for performance of this
surveillance. This would be an exception to the limitations
specified in the Prairie Island Nuclear Generating Plant Technical
Specification Surveillance Requirement 3.0.2 for Surveillance
Requirements with a 24 month Frequency and would allow an interval
up to 30 months for performance of the surveillance.
The changes proposed for the emergency diesel generators do not
change any system operations or maintenance activities. Testing
requirements will be revised and will continue to demonstrate that
the Limiting Conditions for Operation are met and the system
components are functional. The revised test Frequency does not
create new failure modes or mechanisms and no new accident
precursors are generated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This license amendment request proposes to add a Frequency Note
to Surveillance Requirement 3.8.1.8 which will allow application of
the Surveillance Requirement 3.0.2 interval extension (1.25 times
the specified 24 month Frequency) for performance of this
surveillance. This would be an exception to the limitations
specified in the Prairie Island Nuclear Generating Plant Technical
Specification Surveillance Requirement 3.0.2 for Surveillance
Requirements with a 24 month Frequency and would allow an interval
up to 30 months for performance of the surveillance.
The proposed change will continue to ensure that the DG trips
bypass function operates as designed. The functionality and
operability of the emergency power system is not being changed.
Since the requested change only allows extension of the relay
testing interval and failure of the relay by itself does not prevent
the diesel from performing its safety function, this change does not
involve a significant reduction in a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: March 30, 2009.
Description of amendment request: The proposed amendment revises
the Technical Specifications (TS), Appendix A to Facility Operating
License Nos. NPF-2 and NPF-8 for the Joseph M. Farley Nuclear Plant,
Units 1
[[Page 23449]]
and 2, respectively. The changes would eliminate the Reactor Coolant
Pump (RCP) Breaker Position reactor trip. The changes will allow the
elimination of a trip circuitry that is susceptible to single failure
vulnerabilities which can result in unwarranted reactor trips.
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 do not significantly increase the
probability or consequences of an accident previously evaluated in
the Final Safety Analysis Report (FSAR). All of the safety analyses
have been evaluated for impact. The elimination of Reactor Coolant
Pump Breaker Position reactor trip will not initiate any accident;
therefore, the probability of an accident has not been increased. An
evaluation of dose consequences, with respect to the proposed
changes, indicates there is no impact due to the proposed changes
and all acceptance criteria continue to be met. Therefore, these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident than any accident already evaluated in
the FSAR. No new accident scenarios, failure mechanisms or limiting
single failures are introduced as a result of the proposed changes.
The changes have no adverse effects on any safety-related system.
Therefore, all accident analyses criteria continue to be met and
these 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 do not involve a significant reduction in a
margin of safety. All analyses that credit the Reactor Coolant
System Low Flow reactor trip function have been reviewed and no
changes to any inputs are required. The evaluation demonstrated that
all applicable acceptance criteria are met. Therefore, the proposed
changes do not involve a significant reduction in the margin of
safety.
Based on the preceding evaluation, SNC has determined that the
proposed changes meet the requirements of 10 CFR 50.92(c) and do not
involve a significant hazards consideration.
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: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Branch Chief: Melanie C. Wong.
Tennessee Valley Authority (TVA), Docket No. 50 390, Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: April 30, 2009.
Description of amendment request: The proposed amendment would
revise technical specification (TS) Section 5.7, ``Procedures,
Programs, and Manuals,'' to correct typographical errors introduced in
Amendment No. 70, dated October 8, 2008.
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. This change is limited only to correcting a
typographical error in a section number (5.7.2.20 versus 5.2.7.20)
contained in Technical Specification Section 5.0, which will not
change the intent of the added section previously approved in
License Amendment 70. Therefore, no increase in the probability or
consequences of an accident previously evaluated has been created.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. This change is limited only to correcting a
typographical error in a section number (5.7.2.20 versus 5.2.7.20)
contained in Technical Specification Section 5.0, which will not
change the intent of the added section previously approved in
License Amendment 70. Therefore, the possibility of a new or
different kind of accident from those previously analyzed has not
been created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. This change is limited only to correcting a
typographical error in a section number (5.7.2.20 versus 5.2.7.20)
contained in Technical Specification Section 5.0, which will not
change the intent of the added section previously approved in
License Amendment 70. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Based on the above, TVA concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia
Date of amendment request: March 26, 2009.
Description of amendment request: The proposed amendments would
increase each unit's rated thermal power (RTP) level from 2893
megawatts thermal (MWt) to 2940 MWt, and make technical specification
changes as necessary to support operation at the uprated power level.
The proposed change is an increase in RTP of approximately 1.6 percent.
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 will increase the North Anna Power Station
(NAPS) Units 1 and 2 rated thermal power (RTP) from 2893 megawatts
thermal (MWt) to 2940 MWt. Nuclear steam supply systems and balance-
of-plant systems, components and analyses that could be affected by
the proposed change to the RTP were evaluated using revised design
parameters. The evaluations determined that these structures,
systems and components are capable of performing their design
function at the proposed uprated RTP of 2940 MWt. An evaluation of
the accident analyses demonstrates that the applicable analysis
acceptance criteria are still met with the proposed changes. Power
level is an input assumption to equipment design and accident
analyses, but it is not a transient or accident initiator. Accident
initiators are not affected by the power uprate, and plant safety
barrier challenges are not created by the proposed changes.
The radiological consequences of operation at the uprated power
conditions have been assessed. The proposed change to RTP does not
affect release paths, frequency of release, or the analyzed source
term for any accidents
[[Page 23450]]
previously evaluated in the NAPS Updated Final Safety Analysis
Report. Structures, systems and components required to mitigate
transients are capable of performing their design functions with the
proposed changes, and are thus acceptable. Analyses performed to
assess the effects of mass and energy releases remain valid. The
source term used to assess radiological consequences was reviewed
and determined to bound operation at the proposed power level.
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 new accident scenarios, failure mechanisms, or single
failures are introduced as a result of any proposed changes. The UFM
has been analyzed, and system failures will not adversely affect any
safety-related system or any structures, systems or components
required for transient mitigation. Structures, systems and
components previously required for transient mitigation are still
capable of fulfilling their intended design functions. The proposed
changes have no significant adverse affect on any safety-related
structures, systems or components and do not significantly change
the performance or integrity of any safety-related system.
The proposed changes do not adversely affect any current system
interfaces or create any new interfaces that could result in an
accident or malfunction of a different kind than previously
evaluated. Operating at RTP of 2940 MWt does not create any new
accident initiators or precursors. Credible malfunctions are bounded
by the current accident analyses of record or recent evaluations
demonstrating that applicable criteria are still met with the
proposed changes.
Therefore, 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 a
margin of safety?
Response: No.
The margins of safety associated with the power uprate are those
pertaining to core thermal power. These include fuel cladding,
reactor coolant system pressure boundary, and containment barriers.
Core analyses demonstrate that power uprate implementation will
continue to meet the current nuclear design basis. Impacts to
components associated with the reactor coolant system pressure
boundary structural integrity, and factors such as pressure-
temperature limits, vessel fluence, and pressurized thermal shock
were determined to be bounded by the current analyses.
Systems will continue to operate within their design parameters
and remain capable of performing their intended safety functions
following implementation of the proposed change. The current NAPS
safety analyses, including the design basis radiological accident
dose calculations, bound the power uprate.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond,
VA 23219.
NRC Branch Chief: Melanie C. Wong.
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].
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of application for amendment: September 11, 2008, as
supplemented by letter dated December 17, 2008, and January 20, 2009.
Brief Description of amendment: The amendment revised the Technical
Specifications, extending the 15-year interval between containment Type
A tests specified by Specification 4.4.a, ``Integrated Leak Rate
Test,'' by 6 months. The current Type A test interval expires at the
end of April 2009. The amendment extends this interval, on a one-time
basis, to October 2009 to coincide with completion of the next
scheduled refueling outage.
Date of issuance: April 27, 2009.
Effective date: As of the date of issuance and should be
implemented within 60 days.
Amendment No.: 204.
Facility Operating License No. DPR-43: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
65689). The commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 27, 2009.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of application for amendment: July 7, 2008, as supplemented on
September 19, 2008, and March 17, 2009.
Brief description of amendment: The amendment revised the licensing
basis, authorizing the licensee to use the methodology conveyed in the
licensee's letters cited above to determine the seismic loads on the
recently upgraded Auxiliary Building crane. The authorization is
conveyed by addition of a new License Condition 2.C.(11) to Facility
Operating License DPR-43.
Date of issuance: April 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 205.
[[Page 23451]]
Facility Operating License No. DPR-43: The amendment revised
Facility Operating License No. DPR-43.
Date of initial notice in Federal Register: August 26, 2008 (73 FR
50358). The Commission's related evaluation of the amendment is
contained in a safety evaluation dated April 30, 2009.
No Significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: July 16, 2008, as supplemented
by letters dated January 2 and March 19, 2009.
Brief description of amendment: The amendment revised Technical
Specifications 3.1.4, ``Control Rod Scram Times,'' 3.2.2, ``Minimum
Critical Power Ratio (MCPR),'' and 5.6.3, ``Core Operating Limits
Report (COLR),'' to allow incorporation of the analytical methodologies
associated with operation of Global Nuclear Fuel-Americas (GNF) fuel
into the licensing basis to support transition to GNF GE14 fuel.
Date of issuance: May 5, 2009.
Effective date: As of its date of issuance and shall be implemented
prior to beginning operating cycle 20.
Amendment No.: 211.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 14, 2008 (73 FR
60729).
The supplements dated January 2 and March 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 as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 5, 2009.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 18, 2008, as supplemented by
letter dated February 4, 2009.
Brief description of amendment: The amendment modified Technical
Specification (TS) requirements for inoperable snubbers by relocating
the current TS 3.7.8, ``Snubbers,'' to the Technical Requirements
Manual and adding Limiting Condition for Operation (LCO) 3.0.8. The
amendment also made conforming changes to TS LCO 3.0.1. The proposed
amendment is consistent with U.S. Nuclear Regulatory Commission (NRC)-
approved Technical Specification Task Force (TSTF) Improved Standard
Technical Specifications Change Traveler, TSTF-372, Revision 4,
``Addition of LCO 3.0.8, Inoperability of Snubbers,'' as part of the
consolidated line item improvement process.
Date of issuance: May 1, 2009.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 219.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 16, 2008 (73
FR 76410). The supplemental letter dated February 4, 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 May 1, 2009.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: November 18, 2008.
Brief description of amendment: This amendment modifies Technical
Specification 5.5.6 to incorporate Technical Specification Task Force
(TSTF) Travelers TSTF-479, ``Changes to Reflect Revision of 10 CFR
[Code of Federal Regulations] 50.55a,'' and TSTF-497, ``Limit Inservice
Testing Program SR [Surveillance Requirement] 3.0.2 Application to
Frequencies of 2 Years or Less.''
Date of issuance: May 1, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 151.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: January 27, 2009 (74 FR
4772). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 1, 2009.
No significant hazards consideration comments received: No.
GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear
Station, Unit 2, Dauphin County, Pennsylvania
Date of amendment request: June 11, 2008, as supplemented by
letters dated September 15, 2008, December 10, 2008, and March 16,
2009.
Brief description of amendment: The amendment deletes Technical
Specification 6.5, which provided the requirements related to review
and audit functions.
Date of issuance: May 1, 2009.
Effective date: May 1, 2009.
Amendment No.: 63.
Possession Only License No. DPR-73: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 26, 2008 (73 FR
50356) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation Report, dated May 1, 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: June 27, 2007, as supplemented
on April 28, September 4, and December 17, 2008.
Brief description of amendment: The amendment revises surveillance
requirements in Technical Specifications (TS) Section 3.8.1, ``AC
Sources--Operating,'' associated with the diesel generator (DG) steady-
state frequency and voltage. The amendment corrects non-conservative TS
frequency and voltage values, which the licensee states have the
potential to result in undesirable effects such as centrifugal charging
pump motor brake horsepower exceeding its nameplate maximum horsepower,
and subsequently overloading the DGs.
Date of issuance: April 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance April 30, 2009.
Amendment Nos.: 309 (CNP-1), 291 (CNP-2).
Facility Operating License Nos. DPR-58 and DPR-74: Amendment
revises the Renewed Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
[[Page 23452]]
65696). The April 28 and December 17, 2008 supplements provided
additional information that clarified the application, but did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed significant hazards consideration
published in the Federal Register on August 14, 2007.
The September 4, 2008 supplement provided additional information
which expanded the scope of the application as originally noticed. The
NRC staff identified that the specified DG voltage of 3,740 volts at 10
seconds after the DG start was non-conservative and inconsistent with
the 3,910 volt minimum steady-state voltage provided in other parts of
TS Section 3.8.1. The licensee proposed additional changes to TS
Section 3.8.1 in its September 4, 2008 letter. The NRC staff determined
that the proposed expanded scope of the amendment involved a proposed
no significant hazards consideration as published in the Federal
Register on November 4, 2008 (73 FR 65696).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 30, 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: February 24, 2009.
Brief description of amendments: The amendments deleted the
requirement for the power range neutron flux rate-high negative rate
trip (Function 3.b) in Technical Specification (TS) Table 3.3.1-1,
``Reactor Trip System Instrumentation.'' The changes are consistent
with the NRC-approved methodology presented in Westinghouse Topical
Report, WCAP-11394-P-A, ``Methodology for the Analysis of the Dropped
Rod Event,'' dated January 1990. The amendments also incorporated
editorial changes to reflect the deletion of Function 3.b in TS Table
3.3.1-1.
Date of issuance: April 29, 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--205; Unit 2--206.
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: March 24, 2009 (74 FR
12394).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 29, 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: December 31, 2008 superseded the
application dated August 1, 2008, as supplemented by letters dated
November 25 and December 31, 2008.
Brief description of amendment: The amendment revised WBN Unit 1
Technical Specification (TS) 4.2.1, ``Fuel Assemblies,'' and TS
surveillance requirements (SRs) 3.5.1.4, ``Accumulators,'' and 3.5.4.3,
``RWST [Refueling Water Storage Tank],'' to increase the maximum number
of tritium producing burnable absorber rods from 400 to 704.
Date of issuance: April 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment No.: 77.
Facility Operating License No. NPF-90: Amendment revises the TS
4.2.1 and TS SRs 3.5.1.4 and 3.5.4.3.
Date of initial notice in Federal Register: Originally November 12,
2008 (73 FR 66946) was superseded by a notice on January 27, 2009 (74
FR 4776).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 30, 2009.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 7th day of May 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-11268 Filed 5-18-09; 8:45 am]
BILLING CODE 7590-01-P