[Federal Register Volume 74, Number 172 (Tuesday, September 8, 2009)]
[Notices]
[Pages 46239-46247]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-21389]
[[Page 46239]]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0388]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) is publishing this regular biweekly notice. The Act
requires the Commission publish notice of any amendments issued, or
proposed to be issued and grants the Commission the authority to issue
and make immediately effective any amendment to an operating license
upon a determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 13, 2009, to August 26, 2009. The
last biweekly notice was published on August 25, 2009 (74 FR 42926).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking
and Directives Branch (RDB), TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be faxed to the RDB at 301-492-3446. Documents may be examined, and/or
copied for a fee, at the NRC's Public Document Room (PDR), located at
One White Flint North, Public File Area O1F21, 11555 Rockville Pike
(first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to
[[Page 46240]]
participate fully in the conduct of the hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, any hearing held
would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve all adjudicatory documents
over the internet, or in some cases to mail copies on electronic
storage media. Participants may not submit paper copies of their
filings unless they seek an exemption in accordance with the procedures
described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the petitioner/requestor
should contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer\TM\ to
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available
at http://www.nrc.gov/site-help/e-submittals/install-viewer.html.
Information about applying for a digital ID certificate is available on
NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory e-
filing system may seek assistance through the ``Contact Us'' link
located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC Meta-System Help Desk, which is
available between 8 a.m. and 8 p.m., Eastern Time, Monday through
Friday, excluding government holidays. The Meta-System Help Desk can be
contacted by telephone at 1-866-672-7640 or by e-mail at
[email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the request and/
or petition should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii).
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings, unless an NRC regulation or
other law requires submission of such information. With respect to
copyrighted works, except for limited excerpts that serve the purpose
of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submissions.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have
access to ADAMS or who encounter problems in accessing the documents
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].
[[Page 46241]]
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: July 13, 2007, as supplemented July 13,
September 12, November 19, December 13, and December 17, 2007; January
10 (4 letters), January 11 (4 letters), January 14, January 18 (5
letters), January 31, February 25 (2 letters), March 5, and September
30, 2008; March 5 and March 23, 2009.
Description of amendment request: The proposed license amendment
request would revise the Millstone Power Station, Unit No. 3 (MPS3)
spent fuel pool (SFP) storage requirements.
The July 13, 2007 license amendment request proposed a stretch
power uprate (SPU) of MPS3. Included in a supplement dated July 13,
2007, was a request to amend the MPS3 SFP storage requirements. The
July 13, 2007 request was noticed in the Federal Register on January
15, 2008 (73 FR 2549). By letter dated March 5, 2008, Dominion Nuclear
Connecticut, Inc. (DNC) separated the MPS3 SFP storage requirements
request from the MPS3 SPU request.
The request to revise the MPS3 SFP storage requirements is being
re-noticed using the original significant hazards consideration,
specific to the request to revise the SFP storage requirements, as
provided by DNC in the July 13, 2007 license amendment request.
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:
6.1.11.1 [Do the proposed changes] [i]nvolve a significant
increase in the probability or consequences of an accident
previously evaluated[?]
[Response: No]
As discussed in LR [license report] Section 2.8.6.2 [Spent Fuel
Storage] and Westinghouse report WCAP-16721-NP ``Spent Fuel
Criticality Safety Analysis,'' revised spent fuel pool criticality
analyses were performed to take into account the potential for more
reactive fuel at SPU conditions. There are three different regions
defined in the MPS3 spent fuel pool.
Region 1--350 storage locations
Region 2--673 storage locations
Region 3--756 storage locations
Because of the potential for requiring more fresh assemblies to
be loaded in the core every cycle, some of the assemblies to be
discharged to the spent fuel pool may not have sufficient burnup to
meet the requirements of Region 2. It may be necessary to
temporarily store the discharge assemblies in Region 1. To limit the
time that these assemblies need to be stored in Region 1, additional
curves have been added to TS [technical specification] Figure 3.9-3
that specify the burnup limits as a function of enrichment, burnup,
and decay time. These decay time curves provide assurance that all
spent fuel pool criticality limits will be met.
The spent fuel pool criticality analysis also shows that more
limiting burnup requirements are necessary for Region 3 for the
assemblies used at the uprate power level. Thus, a new curve is
being added to address these requirements for Region 3.
With these changes, the spent fuel pool criticality analysis
documented in LR Section 2.8.6.2 and WCAP-16721-NP, shows that the
changes do not increase the consequences of any accident.
The new TS limitations provide assurance that the spent fuel
pool will remain subcritical for all future cycles at the SPU
condition and there is no increase in the probability of a
criticality accident. Thus, the changes do not significantly
increase the probability of any analyzed accident.
6.1.11.2 [Do the proposed changes] [c]reate the possibility of a
new or different kind of accident from any accident previously
evaluated[?]
[Response: No]
The changes will be implemented with existing spent pool racks.
Thus, no new failure modes are introduced. The proposed additional
requirements and the SPU fuel criticality analysis provide assurance
that the spent fuel pool will remain subcritical for all uprate
cycles. Thus, the changes do not create the possibility of a new or
different accident.
6.1.11.3 [Do the proposed changes] [i]nvolve a significant
reduction in a margin of safety[?]
[Response: No]
The analysis documented in LR Section 2.8.6.2 and WCAP-16721-NP
shows that all spent fuel criticality limits are met and that there
is no significant reduction in the margin of safety for the spent
fuel pool.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
NRC Branch Chief: Harold K. Chernoff.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: June 30, 2009.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7.9, ``Ultimate Heat Sink
(UHS),'' to add additional essential service water (SX) cooling tower
fan requirements as a function of SX pump discharge temperature to
reflect the results of a revised analysis for the UHS.
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 does not result in any physical changes to
safety related structures, systems, or components. The UHS itself is
not an accident initiator; rather, the UHS performs functions to
mitigate accidents by serving as the heat sink for safety related
equipment. Consequently, the proposed change does not increase the
probability of occurrence for any accident previously evaluated.
The UHS plays a vital role in mitigating the consequences of any
accident or transient. The proposed changes will ensure that the
minimum conditions necessary for the UHS to perform its design
functions will always be met. Engineering calculations demonstrate
that the SX pump discharge design temperature limit of 100 [deg]F,
which was assumed as an initial input for the accident analyses, is
preserved. Consequently, the proposed changes to cooling tower fan
requirements, relative to the SX pump discharge temperature, do not
increase the consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The supporting analyses for the proposed change do not involve a
new or different kind of accident from any accident previously
evaluated. The proposed limits on maximum SX pump discharge
temperature, and the proposed fan requirements, are within the
design capabilities of the UHS and ensure that the UHS will always
be in a condition to perform its design function in the event of an
accident or transient. New and revised analyses that support the
requested TS changes ensure the full qualification of the UHS. No
changes are being made to the physical design of the UHS such that
the possibility of a new or different kind of accident would be
created. Consequently, these changes do not create the possibility
of a new or different kind of accident from those previously
evaluated.
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?
[[Page 46242]]
Response: No.
The proposed limits on SX pump discharge maximum temperature are
based on the results of new and revised design analyses that ensure
that the margin of safety is not reduced. The new limits on
temperature will ensure that, under the most limiting accident or
transient scenario, cooling water will meet the accident analyses SX
design temperature limit of 100 [deg]F.
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: Stephen J. Campbell.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: March 26, 2009.
Description of amendment request: The proposed amendments would
revise the technical specification (TS) 3.5.1, ``Emergency Core Cooling
System (ECCS) Operating,'' to delete the existing allowance associated
with the automatic depressurization system (ADS) accumulator backup
compressed gas system that currently allows a completion time of 72
hours to restore bottle pressure to >= 500 psig.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not involve a significant increase in
the probability of an accident previously evaluated. The ADS
accumulator backup compressed gas system is designed to maintain the
availability of a mitigation system. It is not recognized as the
initiator of any accident. The failure of the ADS accumulator backup
compressed gas system will not propagate into the onset of an
analyzed event. As such, this proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
This proposed change does not involve a significant increase in
the consequences of an accident previously evaluated. Deleting the
existing allowance associated with the inoperability of the ADS
accumulator backup compressed gas system provides assurance that the
design function of the ADS SRVs [safety relief valves] assumed in
the safety analyses will be achieved under all postulated
conditions. The change that deletes the existing allowable
completion time for an inoperable ADS accumulator backup compressed
gas system is in the conservative direction and will revise the
existing non-conservative TS to be consistent with existing
licensing requirements for multiple inoperable ADS valves.
Therefore, this proposed change will not increase the consequences
of an accident previously evaluated in the UFSAR [updated final
safety analysis report].
Based on the above information, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not affect the control parameters
governing unit operation or the response of plant equipment to
transient conditions. The proposed change does involve the addition
of a reserve nitrogen bottle that can be valved in during bottle
replacement, however, during the short duration the reserve nitrogen
bottle will be valved in the required minimum bottle pressure will
be maintained at 1100 psig. The reserve bottle pressure requirement
for this short duration ensures that the safety function of the ADS
SRVs continues to be met.
Deleting the existing allowance associated with the
inoperability of the ADS accumulator backup compressed gas system
does not introduce any new or different modes of plant operation,
nor does it affect the operational characteristics of any safety-
related equipment or systems; as such, no new failure modes are
being introduced. The proposed action provides assurance that the
design function of the ADS SRVs assumed in the safety analyses will
be achieved; and, therefore the LCO [limiting condition for
operation] will be met. The change that deletes the existing
allowable completion time for an inoperable ADS accumulator backup
compressed gas system is in the conservative direction and will
revise the existing non-conservative TS to be consistent with
existing licensing requirements for multiple inoperable ADS valves.
Based on the above information, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The margin of safety is determined by the design and
qualification of the plant equipment, the operation of the plant
within analyzed limits, and the point at which protective or
mitigative actions are initiated. The modified TS and TRM [Technical
Requirements Manual] will ensure sufficient nitrogen supply exists
to support both the LLS [low-low setpoint] and ADS function of the
SRVs plus assumed design leakage with no operator action.
The change that deletes the existing allowable completion time
for an inoperable ADS accumulator backup compressed gas system is in
the conservative direction and will revise the existing non-
conservative TS to be consistent with existing licensing
requirements for multiple inoperable ADS valves.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: November 6, 2008, as revised by letter
dated August 4, 2009.
Description of amendments request: The proposed change would revise
the Crystal River Unit 3 Improved Technical Specifications Surveillance
Requirements (SRs); SR 3.8.1.2, SR 3.8.1.6, and SR 3.8.1.10 to restrict
the voltage and frequency limits for all emergency diesel generator
(EDG) starts. The steady state voltage limits would be revised to be
more restrictive (plus or minus 2 percent of the nominal voltage) to
accurately reflect the appropriate calculation and the way the plant is
operated and tested. The steady state frequency limits will be revised
to be more restrictive (plus or minus 1 percent for all EDG starts) to
ensure compliance with the plant design bases and the way the plant is
operated. Additionally, SR 3.8.1.6 will be revised to clarify that the
10-second start verifies the capability of the EDG to pick up load, and
is not the steady state condition. These changes will ensure that the
EDGs are capable of supplying power, with the correct voltage and
frequency, to the required electrical loads.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 46243]]
issue of no significant hazards consideration, which is presented
below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The LAR [license amendment request] proposes to provide more
restrictive voltage and frequency limits for the Emergency Diesel
Generators (EDGs) steady state operation. The voltage band is going
from a range of greater than or equal to 3933 V [volts] but less
than or equal to 4400 V, to greater than or equal to 4077 V but less
than or equal to 4243 V. The proposed limits are plus or minus 2%
[percent] around the nominal safety-related bus voltage of 4160 V.
The Frequency Limits are going from a 2% tolerance band to a 1%
tolerance band around the nominal frequency of 60 Hz [hertz] (59.4
Hz to 60.6 Hz) for all starts of the EDGs, at steady state
conditions. For fast starts, the voltage and frequency limits at
less than or equal to ten seconds will be consistent with the EDG
ready matrix setpoints (90.8% voltage and 98% frequency) to allow
for the overshoot and undershoot condition that exists while the
voltage and frequency values converge on steady state conditions.
The EDGs are a safety-related system that functions to mitigate
the impact of an accident with a concurrent loss of offsite power. A
loss of offsite power is typically a significant contributor to
postulated plant risk and, as such, onsite AC generators have to be
maintained available and reliable in the event of a loss of offsite
power event. The EDGs are not initiators for any analyzed accident,
therefore; the probability for an accident that was previously
evaluated is not increased by this change. The revised, voltage and
frequency limits will ensure the EDGs will remain capable of
performing their design function.
The consequences of an accident refer to the impact on both
plant personnel and the public from any radiological release
associated with the accident. The EDG supports equipment that is
supposed to preclude any radiological release. More restrictive
voltage and frequency limits for the output of the EDG restores
design margin, and provides assurance that the equipment supplied by
the EDG will operate correctly and within the assumed timeframe to
perform their mitigating functions.
Until the proposed Crystal River Unit 3 (CR-3) Improved
Technical Specifications (ITS) EDG voltage and frequency limits are
approved by the NRC, administratively controlled limits have been
established in accordance with NRC Administrative Letter 98-10 to
ensure all EDG mitigation functions will be performed, per design,
in the event of a loss of offsite power. These administrative limits
have been determined as acceptable and have been incorporated into
the surveillance test procedures under the provisions of'10 CFR
50.59. Periodic testing has been performed with acceptable results.
Since EDGs are mitigating components and are not initiators for any
analyzed accident, no increased probability of an accident can
occur. Since administrative limits will ensure the EDGs will perform
as designed, consequences will not be significantly affected.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Administrative voltage limits were established using verified
design calculations and the guidance of NRC Administrative Letter
98-10. These administrative limits will ensure the EDGs will perform
as designed. No new configuration is established by this change. The
administrative limits for the EDG frequency were determined to be
sufficient to account for measurement and other uncertainties.
The proposed amendment will place the administrative limits into
the CR-3 ITS. The more restrictive voltage and frequency limits will
provide additional assurance that the EDG can provide the necessary
power to supply the required safety-related loads during an analyzed
accident. The proposed ITS voltage and frequency limits restore the
EDG capability to those analyzed by Engineering calculation. No new
configuration is established. Therefore, no new or different kind of
accident from any previously evaluated can be created.
3. Does not involve a significant reduction in a margin of
safety.
The LAR proposes to provide more restrictive steady state
voltage and frequency limits for the EDGs. The change in the
acceptance criteria for specific surveillance testing provides
assurance that the EDGs will be capable of performing their design
function. Previous test history has shown that the new limits are
well within the capability of the EDGs and are repeatable. The ``as-
left'' settings for voltage and frequency will be adjusted such that
they remain within a tight band and this ensures that the ``as-
found'' settings will be in an acceptable tolerance band.
The proposed ITS limits on voltage and frequency will ensure
that the EDG will be able to perform all design functions assumed in
the accident analyses. Administrative limits are in place to ensure
these parameters remain within analyzed limits. As such, 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: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: June 24, 2009.
Description of amendment request: The proposed amendments would
modify Technical Specification (TS) requirements related to control
room envelope (CRE) habitability in accordance with Technical
Specification Task Force (TSTF) traveler TSTF-448 Revision 3, ``Control
Room Habitability,'' per the consolidated line item improvement process
(CLIIP).
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on October 17, 2006
(71 FR 61075), on possible amendments concerning this CLIIP, including
a model safety evaluation and a model no significant hazards
consideration (NSHC) determination. The NRC staff subsequently issued a
notice of availability of the models for referencing in license
amendment applications in the Federal Register on January 17, 2007 (72
FR 2022), as part of the CLIIP. In its application dated June 24, 2009,
the licensee affirmed the applicability of the following determination.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the
[[Page 46244]]
CRE emergency ventilation system is capable of adequately mitigating
radiological consequences to CRE occupants during accident
conditions, and that the CRE emergency ventilation system will
perform as assumed in the consequence analyses of design basis
accidents. Thus, the consequences of any accident previously
evaluated are not increased. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee and
based on its review, it appears that the standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1
and 2, Will County, Illinois.
Date of amendment request: June 24, 2009.
Brief description of amendment request: The proposed amendment
would permanently revise Technical Specification (TS) 5.5.9, ``Steam
Generator (SG) Program,'' to exclude portions of the tube below the top
of the SG tubesheet from periodic SG tube inspections and plugging or
repair. In addition, this amendment would revise the wording of
reporting requirements in TS 5.6.9, ``Steam Generator (SG) Tube
Inspection Report.'' The proposed changes only affect Byron, Unit No.
2, and Braidwood, Unit 2; however, this action is docketed for both
Byron and Braidwood units because the TS are common to Units 1 and 2.
Date of publication of individual notice in Federal Register: July
31, 2009 (74 FR 38234).
Expiration date of individual notice: August 30, 2009 (public
comment); September 29, 2009 (hearing requests).
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 O1F21, 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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendment: August 29, 2008, as supplemented
by letters dated March 5 and August 7, 2009.
Brief description of amendment: The amendments modified Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' by
updating the list of references in TS 5.6.5.b to reflect the current
analytical methods used to determine the core
[[Page 46245]]
operating limits for Palo Verde Nuclear Generating Station Units 1, 2,
and 3.
Date of issuance: August 26, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 1-174; Unit 2-174; Unit 3-174.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendment revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
65688). The supplemental letters dated March 5 and August 7, 2009,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 26, 2009.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: August 21, 2008.
Brief description of amendments: The amendments implement Technical
Specification Task Force (TSTF) Changes Travelers TSTF-479, Revision 0,
``Changes to Reflect Revision of [Title 10 of the Code of Federal
Regulations] 10 CFR 50.55a,'' and TSTF-497, Revision 0, ``Limit
Inservice Testing [IST] Program SR [Surveillance Requirements] 3.0.2
Application to Frequencies of 2 Years or Less.'' TSTF-479 and TSTF-497
revise the technical specification's Administrative Controls section
pertaining to requirements for the IST Program, consistent with the
requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are
classified as American Society of Mechanical Engineers (ASME), Boiler
and Pressure Vessel Code Class 1, Class 2, and Class 3.
Date of issuance: August 17, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 252 and 232.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: April 3, 2009 (74 FR
18253).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 17, 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: September 9, 2008, as
supplemented by letter dated April 24, 2009.
Brief description of amendment: This amendment modified Technical
Specification 3.3.6.1, ``Primary Containment Isolation
Instrumentation,'' to lower the Group 1 isolation valves reactor water
level isolation signal from Level 2 to Level 1.
Date of issuance: August 18, 2009.
Effective date: As of its date of issuance and shall be implemented
prior to entry into Mode 2 during startup from refueling outage 20.
Amendment No.: 214.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 2, 2008 (73 FR
73353). The supplemental letter dated April 24, 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 August 18, 2009.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: March 5, 2009, as supplemented
by letters dated April 17 and June 22, 2009.
Brief description of amendment: The amendment updates the reactor
vessel heatup and cooldown limit curves and the low-temperature over-
pressure protection curves.
Date of issuance: August 17, 2009.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 262.
Facility Operating License No. DPR-26: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: May 19, 2009 (74 FR
23443). The April 17 and June 22, 2009, supplements provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 17, 2009.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: April 16, 2009.
Description of amendment request: This amendment changes the name
of the Licensee and Co-owner from ``FPL Energy Seabrook, LLC'' to
``NextEra Energy Seabrook, LLC.''
Date of issuance: August 21, 2009.
Effective date: As of its date of issuance and shall be implemented
within 30 days.
Amendment No.: 122.
Facility Operating License No. NPF-86: The amendment revised the
License, Appendix B--Environmental Protection Plan, and Appendix C--
Additional Conditions.
Date of initial notice in Federal Register: June 2, 2009 (74 FR
26434).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 21, 2009.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2 (NMP2), Oswego County, New York
Date of application for amendment: March 9, 2009.
Brief description of amendment: The amendment revises the Technical
Specification (TS) testing frequency for the Surveillance Requirement
(SR) in TS 3.1.4, ``Control Rod Scram Times,'' by extending the
frequency of SR 3.1.4.2, from ``120 days cumulative operation in Mode
1'' to ``200 days cumulative operation in Mode 1.'' This change is
based on Nuclear Regulatory Commission-approved TS Task Force (TSTF)
Change Traveler, TSTF-460-A, Revision 0, ``Control Rod Scram Time
Testing Frequency.'' These changes were described in a Notice of
Availability for Consolidated Line Item
[[Page 46246]]
Improvement Process published in the Federal Register on August 23,
2004 (69 FR 51864).
Date of issuance: August 19, 2009.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 132.
Renewed Facility Operating License No. NPF-069: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: May 19, 2009 (74 FR
23447).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 19, 2009.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, LLC, Docket No. 50-263,
Monticello Nuclear Generating Plant, Wright County, Minnesota
Date of application for amendment: April 15, 2009.
Brief description of amendment: The amendment revised the MNGP
Technical Specifications (TS), deleting paragraph d (regarding
limitation of working hours of personnel who perform safety-related
functions) of TS 5.2.2, ``Unit Staff.''
Date of issuance: August 19, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009.
Amendment No.: 163.
Facility Operating License No. DPR-22. Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 16, 2009 (74 FR
28578).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 19, 2009.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of application for amendments: April 15, 2009.
Brief description of amendments: The amendments delete those
portions of the Technical Specifications superseded by Title 10 of the
Code of Federal Regulations Part 26, Subpart I.
Date of issuance: August 19, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009.
Amendment Nos.: 193, 182.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 16, 2009 (74 FR
28578).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 19, 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 (DCPP), San Luis Obispo
County, California
Date of application for amendments: May 5, 2009.
Brief description of amendments: The amendments revised the DCPP
Technical Specification (TS) 5.2.2, ``Unit Staff,'' to eliminate
working hour restrictions in paragraph d of TS 5.2.2 to support
compliance with Title 10 of the Code of Federal Regulations (10 CFR)
Part 26. The change is consistent with U.S. Nuclear Regulatory
Commission (NRC)-approved Revision 0 to TS 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 availability of this TS improvement was announced in the
Federal Register on December 30, 2008 (73 FR 79923), as part of the
consolidated line item improvement process.
Date of issuance: August 19, 2009.
Effective date: As of its date of issuance and shall be implemented
by October 1, 2009.
Amendment Nos.: Unit 1-206; Unit 2-207.
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: June 16, 2009 (74 FR
28579).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 19, 2009.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: December 4, 2008.
Brief description of amendment: The amendment revises the Technical
Specifications to allow refueling operations with both containment
personnel interlock doors to be open under administrative control
consistent with Technical Specification Task Force (TSTF) Travelers
TSTF-68 and TSTF-312. In support of this amendment request, the
licensee recalculated the fuel gas gap fractions for its design-basis
fuel handling accident and has justified a shorter decay time of 72
hours utilizing the alternative source term methodology.
Date of issuance: August 12, 2009
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 107
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: March 10, 2009 (74 FR
10311)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 12, 2009.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: March 23, 2009.
Brief description of amendment: The amendment deletes paragraph d
of Technical Specification (TS) 5.2.2, ``Plant Staff.'' The amendment
is consistent with Nuclear Regulatory Commission approved Revision 0 to
the Technical Specification Task Force (TSTF) Improved Standard
Technical Specification Change Traveler, TSTF-511, ``Eliminate Working
Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR
[Title 10 of the Code of Federal Regulations] Part 26.'' The
availability of this TS improvement was announced in the Federal
Register on December 30, 2008 (73 FR 79923) as part of the consolidated
line item improvement process.
Date of issuance: August 12, 2009.
Effective date: As of the date of issuance to be implemented with
the implementation of the new 10 CFR Part 26, Subpart I requirements.
Amendment No.: 108.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: April 21, 2009 (74 FR
18256).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 12, 2009.
No significant hazards consideration comments received: No.
[[Page 46247]]
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 3, 2009.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) to eliminate working hour restrictions
from TS 6.2.2 to support compliance with Title 10 of the Code of
Federal Regulations (10 CFR) Part 26. The request is consistent with
the guidance contained in the U.S. Nuclear Regulatory Commission (NRC)-
approved TS Task Force (TSTF) Improved Standard Technical Specification
change traveler, TSTF-511, Revision 0, ``Eliminate Working Hour
Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.''
The availability of this improvement was announced in the Federal
Register on December 30, 2008 (73 FR 79923), as part of the
Consolidated Line Item Improvement Process.
Date of issuance: August 18, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009.
Amendment Nos.: Unit 1-192; Unit 2-180.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: June 16, 2009 (74 FR
28579).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 18, 2009.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: March 27, 2008, as supplemented
by letters dated December 19, 2008, February 9, April 24, and May 26,
2009.
Description of amendment request: The amendments revised the
technical specifications (TSs) to adopt the content of Technical
Specification Task Force (TSTF) change traveler TSTF448, Revision 3,
``Control Room Habitability.'' Specifically, the amendments revised TS
3.7.3, ``Control Room Emergency Ventilation (CREV) System,'' and added
TS 5.5.13, ``Control Room Envelope Habitability Program.'' The
amendments also added a new license condition regarding initial
performance of the new surveillance and assessment requirements of the
revised TSs.
Date of issuance: August 18, 2009.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 275, 302, and 261.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Licenses and Technical Specifications.
Date of initial notice in Federal Register: August 26, 2008 (73 FR
50362) and revised on January 27, 2009 (74 FR 4775). The supplements
dated February 9, April 24, and May 26, 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 August 18, 2009.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: October 21, 2008.
Brief description of amendments: The amendments revised Sequoyah
Nuclear Plant's Updated Final Safety Analysis Report (UFSAR) to require
an inspection of each ice condenser within 24 hours of experiencing a
seismic event greater than or equal to an operating basis earthquake
(i.e., \1/2\ of a safe shutdown earthquake) within the 5-week period
after ice basket replenishment is completed. This will confirm that ice
condenser lower inlet doors have not been blocked by ice fallout.
The proposed amendments provided a procedural requirement to
confirm the ice condenser maintains the ice condenser generic
qualification as set forth in the UFSAR. Justification for the use of
the proposed procedural requirement is based on reasonable assurance
that the ice condenser lower inlet doors will open following a seismic
event during the 5-week period and the low probability of a seismic
event occurring coincident with or subsequently followed by a design
basis accident.
Date of issuance: August 14, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance. The UFSAR changes shall be implemented in
the next periodic update made in accordance with 10 CFR 50.71(e).
Amendment Nos.: 325 and 317.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
changed the licenses.
Date of initial notice in Federal Register: January 13, 2009 (74 FR
1715).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 14, 2009.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 27th day of August, 2009.
For The Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-21389 Filed 9-4-09; 8:45 am]
BILLING CODE 7590-01-P