[Federal Register Volume 75, Number 229 (Tuesday, November 30, 2010)]
[Notices]
[Pages 74091-74099]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-29941]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0367]
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 (NRC or the
Commission) 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 November 3, 2010, to November 17, 2010. The
last biweekly notice was published on November 16, 2010 (75 FR 70032).
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), 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, Rules,
Announcements and Directives Branch (RADB), 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 RADB 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, Room O1-F21,
11555 Rockville Pike (first floor), Rockville, Maryland 20852.
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, Room O1-F21,
11555 Rockville Pike (first floor), Rockville, Maryland 20854. Publicly
available records will be accessible from the Agencywide
[[Page 74092]]
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 requestor/petitioner
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
requestor/petitioner 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 requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner 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 requestor/petitioner to relief. A requestor/
petitioner 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, 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 (72 FR 49139,
August 28, 2007). 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 participant should
contact the Office of the Secretary by e-mail at
[email protected], or by telephone at 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
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
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. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant 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
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing 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
E-Filing 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
[[Page 74093]]
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 by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
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. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
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, or the 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.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
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].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: September 24, 2010.
Description of amendment request: The proposed amendment would
revise the Fermi 2 Radiological Emergency Response Preparedness (RERP)
Plan to increase the staff augmentation times for Technical Support
Center-related functions from 30 to 60 minutes and for Emergency
Operations Facility-related functions from 60 to 90 minutes.
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 change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed extension of staff augmentation times has no effect
on normal plant operation or on any accident initiator. The change
affects the response to radiological emergencies under the Fermi 2
Radiological Emergency Response Preparedness (RERP) Plan. The
ability of the emergency response organization to respond adequately
to radiological emergencies has been evaluated. Improvements have
been made to equipment, procedures, and training since initial
approval of the Fermi 2 Emergency Plan that have resulted in a
significant increase in the on-shift capabilities and knowledge such
that there would be no degradation or loss of Emergency Plan
function as a result of the proposed change. A functional analysis
was also performed on the effect of the proposed change on the
timeliness of performing major tasks for the major functional areas
of the RERP Plan. The analysis concluded that extension of staff
augmentation times would not significantly affect the ability to
perform the required tasks.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
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 affects the required response times for
supplementing onsite personnel in response to a Radiological
emergency. It has been evaluated and determined not to significantly
affect the ability to perform that function. It has no effect on the
plant design or on the normal operation of the plant and does not
affect how the plant is physically operated under emergency
conditions. The extension of staff augmentation times in the RERP
Plan does not affect the plant Operating, Abnormal Operating, or
Emergency Operating procedures which are performed by plant staff
during all plant conditions.
Therefore, since the proposed change does not affect the design
or method of operation of the plant, it does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change does not affect plant design or method of
operation. 10 CFR 50.47(b) and 10 CFR Part 50, Appendix E establish
emergency planning standards that require adequate staffing,
satisfactory performance of key functional areas and critical tasks;
and timely augmentation of the response capability. Since the
initial NRC approval of the Emergency Plan, there have been
improvements in the technology used to support the RERP functions
and in the capabilities of onsite personnel. A functional analysis
was performed on the effect of the proposed change on the timeliness
of performing major tasks for the functional areas of the RERP Plan.
The analysis concluded that an increase in staff augmentation times
would not significantly affect the ability to perform the required
RERP tasks. Thus, the proposed change has been determined not to
adversely affect the ability to meet the emergency planning
standards as described in 10 CFR 50.47(b) and 10 CFR Part 50,
Appendix E.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 74094]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David G. Pettinari, Attorney--Corporate
Matters, 688 WCB, Detroit Edison Company, One Energy Plaza, Detroit,
Michigan 48226-1279.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of amendment request: January 6, 2010, as supplemented by
letters dated August 20, 2010, and October 14, 2010.
Description of amendment request: The proposed amendment would
enable PBAPS, Units 2 and 3, to possess byproduct and special nuclear
material from Limerick Generating Station (LGS), Units 1 and 2.
Specifically, the revised license paragraph would permit storage of
low-level radioactive waste (LLRW) from LGS in the PBAPS LLRW Storage
Facility. The PBAPS LLRW Storage Facility currently provides storage
for LLRW generated at PBAPS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (NSHC), 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 is an amendment to the PBAPS FOLs [Facility
Operating Licenses] that will enable PBAPS to receive and store
Class B/C LLRW from LGS in the PBAPS LLRWSF [Low Level Radioactive
Waste Storage Facility]. This proposed change does not impact any
initiators or precursors of previously analyzed accidents. The
storage of Class B/C LLRW from LGS does not impact the failure of
any plant structures, systems, or components. The proposed change
does not have a detrimental impact on the integrity of any plant
structure, system, or component that initiates an analyzed event.
The proposed change does not affect any active or passive failure
mechanisms that could lead to an accident. The PBAPS LLRWSF is not
safety related, and is not used for plant shutdown resulting from
accident or nonstandard operational conditions.
The proposed change does not significantly increase the
consequences of postulated design basis events (i.e., seismic,
flood, tornado, fire, and container drop events), in that the
postulated impact of these events remains well below regulatory
requirements (i.e., less than 10 percent of 10 CFR Part 100,
``Reactor Site Criteria'' acceptance criteria).
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 previously evaluated?
Response: No.
The proposed change is an amendment to the PBAPS FOLs that will
enable PBAPS to receive and store Class B/C LLRW from LGS in the
PBAPS LLRWSF. The proposed amendment does not involve any change to
the plant equipment or system design functions. EGC has verified
that the storage of Class B/C LLRW from LGS in the PBAPS LLRWSF does
not affect the ability of the PBAPS LLRWSF to perform its design
function, including compliance with NRC regulatory requirements and
guidance. No new accident initiators are introduced by this
amendment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The proposed change is an amendment to the PBAPS FOLs that will
enable PBAPS to receive and store Class B/C LLRW from LGS in the
PBAPS LLRWSF. The proposed amendment does not involve any change to
plant equipment or system design functions. The margin of safety is
established through the design of the plant structures, systems, and
components, the parameters within which the plant is operated, and
the setpoints for the actuation of equipment relied upon to respond
to an event. The proposed amendment does not affect the PBAPS safety
limits or setpoints at which protective actions are initiated.
The proposed amendment does not significantly increase the dose
rate at the exterior wall of the LLRWSF, the nearest restricted area
boundary, and the nearest residence when the LLRWSF is filled to
capacity with Class B/C LLRW. Therefore, these dose rates will
remain within limits specified in 10 CFR Part 20 and 40 CFR Part
190.
Additionally, the potential radiological impact of a postulated
design basis container drop accident is less than 10 percent of the
10 CFR Part 100 acceptance criteria.
Therefore the margin of safety is not reduced by the proposed
change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves NSHC.
Attorney for licensee: Mr. J. Bradley Fewell, Associate General
Counsel, Exelon Generation Company LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania.
Date of amendment request: June 25, 2010, as supplemented by letter
dated August 16, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirement (SR)
3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' and SR
3.6.1.5, ``Reactor Building-to-Suppression Chamber Vacuum Breakers,''
to modify the required level for the liquid nitrogen storage tank.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (NSHC), 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 TS changes to increase the level in the liquid
nitrogen storage tank from >= 16 inches water column to a level of
>= 22 inches water column, or equivalent volume of >= 124,000 scf
[standard cubic feet] at 250 psig, is necessary in order to correct
a non-conservative TS value. Increasing the level is intended to
ensure continued operability of the PCIVs (SR 3.6.1.3.1) and Reactor
Building-to-Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) via
the SGIG [safety grade instrument gas] system. The non-conservative
TS condition was identified based on a re-analysis of the liquid
nitrogen storage tank operation. The leakage allowance that was
previously assumed was not based on a rigorous empirical value. The
re-analysis of the leakage allowance assumes more reasonable system
leakage based on operational data. Exelon determined that the
current PBAPS, Units 2 and 3, TS SR value for the minimum level in
the liquid nitrogen storage tank of >= 16 inches water column is
non-conservative and that the guidance of Nuclear Regulatory
Commission (NRC) Administrative Letter 98-10, ``Dispositioning of
Technical Specifications that are Insufficient to Assure Plant
Safety,'' applies. Exelon has implemented administrative controls to
maintain the amount of nitrogen in the liquid nitrogen storage tank
at a level of > 22 inches water column in support of SGIG system
operation.
Exelon is submitting this License Amendment Request to address
this non-conservative condition. The proposed TS
[[Page 74095]]
changes do not introduce new equipment or new equipment operating
modes, nor do the proposed changes alter existing system
relationships. The proposed changes do not affect plant operation,
design function or any analysis that verifies the capability of a
system, structure or component (SSC) to perform a design function.
Further, the proposed changes do not increase the likelihood of the
malfunction of any SSC or impact any analyzed accident.
Consequently, the probability or consequences of an accident
previously evaluated are not affected.
Therefore, the proposed amendments 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 proposed TS change to increase the level in the liquid
nitrogen storage tank from >= 16 inches water column to a level of
>= 22 inches water column, or equivalent volume of >= 124,000 scf at
250 psig, for the PCIVs (SR 3.6.1.3.1) and Reactor Building-to-
Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) is needed to
correct a non-conservative value based on a revised analysis. The
proposed TS changes do not alter the design function or operation of
any SSC. There is no new system component being installed, no
construction of a new facility, and no performance of a new test or
maintenance function. The proposed TS changes do not create the
possibility of a new credible failure mechanism or malfunction. The
proposed changes do not modify the design function or operation of
any SSC. Further, the proposed changes do not introduce new accident
initiators. Consequently, the proposed changes cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Therefore, the proposed amendments do not create the possibility
of a new or different kind of accident from any accident previously
analyzed.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed TS changes to increase the level in the liquid
nitrogen storage tank from >= 16 inches water column to a level of
>= 22 inches water column, or equivalent volume of >= 124,000 scf at
250 psig, for the PCIVs (SR 3.6.1.3.1) and Reactor Building-to-
Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) are necessary to
correct an existing non-conservative TS value. The proposed TS
changes are needed based on a revised analysis that utilizes
empirical data for nitrogen system uses and losses. The proposed
changes do not exceed or alter a design basis or a safety limit for
a parameter established in the PBAPS, Units 2 and 3, Updated Final
Safety Analysis Report (UFSAR) or the PBAPS, Units 2 and 3, Renewed
Facility Operating License (FOL). Consequently, the proposed changes
do not result in a reduction in the margin of safety.
Therefore, the proposed amendments 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 NSHC.
Attorney for licensee: Mr. J. Bradley Fewell, Associate General
Counsel, Exelon Generation Company LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: September 22, 2010.
Description of amendment request: The proposed amendment would
relocate the list of pumps, fans, and valves in Technical Specification
(TS) 4.5.1.1b, Sequence and Power Transfer Test, to the Three Mile
Island, Unit 1 (TMI-1) Updated Final Safety Analysis Report. In
addition, TS 4.5.1.2b, TS 4.5.2.2a, and TS 4.5.2.2b refer to this test
and are proposed for revision to reflect the proposed change to TS
4.5.1.1b.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with an NRC edit in brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed license amendment does not add, delete or modify
plant equipment. The proposed changes are administrative in nature.
The proposed amendment would relocate the list of pumps, fans and
valves in Technical Specification (TS) 4.5.1.1b, Sequence and Power
Transfer Test, to the TMI-1 Updated Final Safety Analysis Report
(UFSAR) Section 8.3, Tests and Inspections.
The proposed changes relocate surveillance requirement details
that are not required by 10 CFR 50.36, and are [partially]
consistent with standard technical specifications, NUREG-1430,
``Standard Technical Specifications Babcock and Wilcox Plants.'' The
proposed changes do not change current surveillance requirements.
The subject list of pumps, fans and valves that will be relocated to
the UFSAR Section 8.3 will continue to be administratively
controlled and future changes will be controlled under 10 CFR 50.59.
The probability of an accident is not increased by these
proposed changes because the Sequence and Power Transfer Test is not
an initiator of any design basis event. Additionally, the proposed
changes do not involve any physical changes to plant structures,
systems, or components (SSCs), or the manner in which these SSCs are
operated, maintained, or controlled. The consequences of an accident
will not be increased because the proposed administrative changes to
the Sequence and Power Transfer Test and Sequence Test will continue
to provide a high degree of assurance that the Electric Power System
will meet its safety related function.
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 previously evaluated?
Response: No.
The proposed changes do not alter the physical design, safety
limits, safety analyses assumptions, or the manner in which the
plant is operated or tested. The proposed changes are administrative
in nature and the surveillance requirements remain the same.
Accordingly, the proposed changes do not introduce any new accident
initiators, nor do they reduce or adversely affect the capabilities
of any plant SSC in the performance of their safety function.
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 the margin of safety?
Response: No.
The margin of safety is associated with the confidence in the
ability of the fission product barriers (i.e., fuel cladding,
reactor coolant pressure boundary, and containment structure) to
limit the level of radiation to the public. There are no physical
changes to SSCs or operating and testing procedures associated with
the proposed amendment.
The proposed changes do not impact the assumptions of any design
basis accident, and do not alter assumptions relative to the
mitigation of an accident or transient event. The proposed changes
are administrative in nature and the surveillance requirements
remain the same.
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, with the NRC edit noted above incorporated, it appears
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
[[Page 74096]]
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: September 24, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.1.2.3, to allow up to two Main
Steam Safety Valves (MSSVs) per steam generator to be inoperable with
no required reduction in power level. It would also revise the required
maximum overpower trip setpoints for any additional inoperable MSSVs
consistent with the plant transient analysis. The proposed change
requires that with less than four MSSVs associated with either steam
generator operable, the plant would be required to be brought to the
hot shutdown condition.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below, with NRC edits in brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed amendment is not a change to the plant structures,
systems, or components. There is no increase to the likelihood of
Main Steam Safety Valve (MSSV) related failures. The MSSVs are
relied upon to mitigate the effects of Updated Final Safety Analysis
Report (UFSAR) Chapter 14 design basis events including the loss of
load (turbine trip), which is the limiting event for secondary
system overpressure. Analyses, performed in accordance with NRC
approved methods, have demonstrated that with reduced MSSV
availability and following the specified power level restrictions,
the MSSVs will continue to limit the secondary system pressure to
less than 110 percent of the design pressure of the Once Through
Steam Generators (OTSGs) and the Main Steam (MS) System as required
by [the American Society of Mechanical Engineers] ASME code.
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 previously evaluated?
Response: No.
The proposed amendment is not a change to the plant structures,
systems, or components (SSCs). Furthermore, within the current
licensing basis, the MSSVs are accident mitigation SSCs. The current
licensing basis does not [explicitly] include consideration of a
MSSV failure as an event initiator [and a failed open MSSV has been
shown to be bounded by the larger maximum break size analysis
presented in the TMI-1 UFSAR]. The proposed amendment will not
fundamentally alter or create any new operator actions. 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 the margin of safety?
Response: No.
The limiting event for secondary system overpressure is a loss
of load event (turbine trip). The event has been analyzed for
varying MSSVs out of service, using NRC approved methods. The
results of the analysis demonstrate that the existing design
acceptance criteria (i.e., MS and OTSG pressure remain less than 110
percent of the design pressure) are met for all combinations of
inoperable MSSVs and initial power levels described in the proposed
change. 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, including the edits listed above, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: January 27, 2010.
Description of amendment request: The proposed changes would amend
Renewed Facility Operating Licenses DPR-24 and DPR-27 for the Point
Beach Nuclear Plant, Units 1 and 2, respectively. The proposed
amendment consists of changes to Technical Specification 3.8.3,
``Diesel Fuel Oil and Starting Air.''
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 increases the minimum required amount of
stored diesel fuel in the associated common fuel oil storage tank
[FOST] for two standby emergency power sources to start, load to
their respective loading limits and to operate continuously up to a
maximum of 48 hours. An increase in the minimum required fuel oil
volume required in the fuel oil storage tanks does not increase the
probability or consequences of an accident previously evaluated.
[Limiting Condition for Operation] LCO 3.8.3 Condition A,
currently requires that one or more standby emergency power sources
have >= 11,000 gallons of fuel when the associated [emergency diesel
generator] EDG is declared operable. The proposed change increases
the amount of stored fuel to >= 24,000 gallons for two standby EDGs.
It further adds new Required Action A.2 if the FOST stored capacity
falls below the minimum required values. The proposed change also
accounts for instrument indicator loop uncertainty values for
unusable volume.
New LCO [3.8.3] Condition B, addresses the case of one EDG
operating in either Train ``A'' or Train ``B.'' The new condition
specifies that the minimum volume of diesel fuel required to support
continued operation of a single EDG for 48 hours at rated load is >=
13,000 gallons. This proposed change also accounts for instrument
indicator loop uncertainty values for unusable volume.
[Surveillance Requirement] SR 3.8.3.1 is revised to reflect the
increased amount of diesel fuel required to be maintained to support
operation of the EDGs following recalculation of required values.
Following implementation of this proposed change, there will be
no change in the ability of the EDGs to supply maximum post-accident
load demands for 48 hours. The proposed minimum volume of fuel, >=
24,000 gallons for two EDGs and >= 13,000 gallons for one EDG per
train, ensures that a 48-hour supply of fuel is available when the
associated standby emergency power source is 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 EDGs and the associated support systems, such as the fuel
oil storage and transfer systems, are designed to mitigate accidents
and are not accident initiators. Following this change, the EDGs
will continue to supply the required maximum post-accident load
demand. The current 48-hour fuel supply requirements will be
maintained following this change. The new required fuel oil volumes
are within the capacities of the fuel oil storage tanks.
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.
There are two underground fuel oil storage tanks on site. Each
tank has a capacity of approximately 35,000 gallons and each common
fuel tank supports one EDG train.
[[Page 74097]]
Fuel can be manually transferred from one tank to another via a
cross-connect valve. Sufficient fuel is maintained between the two
tanks to allow one EDG to operate continuously at the required load
for seven (7) days. At the proposed minimum required level, which is
>= 24,000 gallons in the common fuel oil storage tanks for two
standby emergency power sources, one tank could provide enough fuel
for two EDGs in either Train A or Train B to continue operation for
great than 48 hours. At the proposed minimum required level, which
is >= 13,000 gallons in each fuel oil storage tanks, one tank could
provide enough fuel for one EDG in Train A and Train B to continue
operation for greater than 48 hours.
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: William Blair, Senior Attorney, NextEra
Energy Point Beach, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Robert J. Pascarelli.
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 amendment request: August 27, 2010.
Description of amendment request: The proposed amendment would add
a new Action to Technical Specification (TS) 3.7.3, ``Control Room
Emergency Ventilation (CREV) System,'' to permit one or more CREV
subsystems to be inoperable for up to 90 days when the inoperability is
due to inoperable CREV System High Efficiency Particulate Air (HEPA)
filter and/or charcoal absorbers. The proposed TS changes also include
an administrative change to correct errors in Unit 2 TS page header
information that occurred during issuance of TS pages for a previous
amendment.
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.
UFSAR [Updated Finale Safety Analysis Report] Chapter 14,
``Plant Safety Analysis,'' evaluates operational transients and
accidents that result in radiological releases that affect control
room occupants. UFSAR section 14.6, ``Analysis of Design Basis
Accidents--Uprated,'' evaluates accidents that release fission
products to the environment. The CREV System is not an accident
initiator for any of the accidents described. The CREV System
processes outside air needed to provide ventilation and
pressurization for control room habitability to limit the control
room dose during accidents evaluated in the UFSAR. Without crediting
the performance of the HEPA filter or charcoal adsorbers, the
analyses results concludes that the 30[-]day integrated post-
accident doses in the control room are within the limits of 5 rem
TEDE [total effective dose equivalent], as specified in 10 CFR 50.67
and GDC [General Design Criterion]-19. The control room dose
increase is less than 10 percent; leaving more than 60 percent
remaining margin to the regulatory limit.
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 CREV System is a ventilation system that filters outside air
used to pressurize the control rooms to provide a protected
environment from which operators can control the unit during
airborne challenges from radioactivity during accident conditions.
The CREV System does not initiate accidents. The proposed amendment
allows the CREV HEPA filters and charcoal adsorbers to be repaired
or replaced without shutting down the operating unit(s). No new
modes of operation are introduced.
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.
Analyses associated with the prior approval of Alternate Source
Term methodology for design basis accident dose consequences
previously did not credit the CREV System charcoal adsorbers. Recent
analyses have been performed to assess the post-accident 30-day
control room dose removing credit for the CREV System HEPA filter.
The results indicate a minimal increase in dose consequences (9.5
percent increase) due to removing credit for the CREV System HEPA
filter. Even with no credit for either the CREV System HEPA filter
or CREV System charcoal filter, the resultant control room dose
maintains more than 60 percent margin to the regulatory limit of 5
rem TEDE. As such there is no reduction in a margin of safety for
any duration of inoperability of the CREV System HEPA filter or
charcoal adsorbers. While the HEPA filter and charcoal adsorbers are
not credited for accident mitigation, they remain required by the
BFN TS for compliance with the LCO 3.7.3, ``Control Room Emergency
Ventilation (CREV) System,'' further minimizing any potential
reduction in a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
NRC Branch Chief: Douglas A. Broaddus.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: October 21, 2010.
Description of amendment request: The proposed action involves the
inclusion of the Westinghouse Best-Estimate (BE) Large Break Loss-of-
Coolant Accident (LBLOCA) analysis methodology using the Automated
Statistical Treatment of Uncertainty Method (ASTRUM) for the analysis
of LBLOCA to the list of methodologies approved for reference in the
Core Operating Limits Report (COLR) in Technical Specification (TS)
5.6.5.b. This action also removes four obsolete COLR references that
supported North Anna Improved Fuel (NAIF) product, Westinghouse Vantage
5, since this product is not planned to be used in future North Anna
cores.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[Criterion 1]
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
No physical plant changes are being made as a result of using
the Westinghouse Best Estimate Large Break LOCA (BE-LBLOCA) analysis
methodology. The proposed TS change simply involves updating the
references in TS 5.6.5.b, Core Operating Limits Report (COLR), to
reference the Westinghouse BE-LBLOCA analysis methodology, which is
an NRC approved methodology, and to delete unnecessary references.
Therefore, the probability of LOCA occurrence is not affected by the
change. Further, the consequences of a LOCA are not increased, since
the BE-LBLOCA analysis has demonstrated that the performance of the
Emergency Core Cooling
[[Page 74098]]
System (ECCS) continues to conform to the criteria contained in 10
CFR 50.46, ``Acceptance Criteria for Emergency Core Cooling Systems
for Light-Water Nuclear Power Reactors.'' No other accident
consequence is potentially affected by this change.
Systems will continue to be operated in accordance with current
design requirements under the new analysis, therefore no new
components or system interactions have been identified that could
lead to an increase in the probability of any accident previously
evaluated in the Updated Final Safety Analysis Report (UFSAR). No
changes were required to the Reactor Protection System (RPS) or
Engineering Safety Features (ESF) setpoints because of the new
analysis methodology.
An analysis of the LBLOCA accident for North Anna Units 1 and 2
has been performed with the Westinghouse BE-LBLOCA analysis
methodology using ASTRUM. The analysis was performed in compliance
with the NRC conditions and limitations as identified in WCAP-1
6009- P-A. Based on the analysis results, it is concluded that the
North Anna Units 1 and 2 continue to satisfy the limits prescribed
by 10 CFR 50.46.
There are no changes to assumptions of the radiological dose
calculations. Hence, there is no increase in the predicted
radiological consequences of accidents postulated in the UFSAR.
Therefore, neither the probability of occurrence nor the
consequences of an accident previously evaluated is significantly
increased.
[Criterion 2]
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The use of the Westinghouse BE-LBLOCA analysis methodology with
ASTRUM does not impact any of the applicable design criteria and
pertinent licensing basis criteria continue to be met. Demonstrated
adherence to the criteria in 10 CFR 50.46 precludes new challenges
to components and systems that could introduce a new type of
accident. Safety analysis evaluations have demonstrated that the use
of Westinghouse BE-LBLOCA analysis methodology with ASTRUM is
acceptable. Design and performance criteria continue to be met and
no new single failure mechanisms have been created. The use of the
Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not
involve any alteration to plant equipment or procedures that would
introduce any new or unique operational modes or accident
precursors. Furthermore, no changes have been made to any RPS or ESF
actuation setpoints. Based on this review, it is concluded that no
new accident scenarios, failure mechanisms, or limiting single
failures are introduced as a result of the proposed changes.
Therefore, the possibility for a new or different kind of
accident from any accident previously evaluated is not created.
[Criterion 3]
Does this change involve a significant reduction in a margin of
safety?
Response: No.
It has been demonstrated that the analytical technique used in
the Westinghouse BE-LBLOCA analysis methodology using ASTRUM
realistically describes the expected behavior of the reactor system
during a postulated LOCA. Uncertainties have been accounted for as
required by 10 CFR 50.46. A sufficient number of LOCAs with
different break sizes, different locations, and other variations in
properties have been considered to provide assurance that the most
severe postulated LOCAs have been evaluated. The analysis has
demonstrated that the acceptance criteria contained in 10 CFR 50.46
continue to be satisfied.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: Gloria Kulesa.
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, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20854. Publicly available records will be accessible from the
Agencywide Documents Access and Management System (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: April 29, 2010.
Brief description of amendment: The amendments adopted Nuclear
Regulatory Commission (NRC)-approved TS Task Force (TSTF) Standard
Technical Specification change traveler TSTF-491, Revision 2, ``Removal
of Main Steam and Main Feedwater Valve Isolation Times from Technical
Specifications.'' The isolation times will be located outside of the
TSs in a document subject to control by the 10 CFR 50.59 process.
Date of issuance: November 5, 2010.
Effective Date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 1--181; Unit 2--181; Unit 3--181.
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 the Federal Register: July 27, 2010 (75
FR 44024).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 5, 2010.
No significant hazards consideration comments received: No.
[[Page 74099]]
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: June 23, 2010.
Brief description of amendment: Current Technical Specification
(TS) 6.5.8, ``Inservice Testing Program,'' contains references to the
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code, Section XI as the source of requirements for the inservice
testing (IST) of ASME Code Class 1, 2, and 3 pumps and valves. The
amendment deleted the references to Section XI of the Code and
incorporated references to the ASME Code for Operation and Maintenance
of Nuclear Power Plants (ASME OM Code). The amendment also indicates
that there may be some nonstandard frequencies utilized in the IST
Program in which the provisions of Surveillance Requirement (SR) 3.0.2
are applicable. The changes are consistent with Technical Specification
Task Force (TSTF) Technical Change Travelers TSTF-479-A, ``Changes to
Reflect Revision of 10 CFR 50.55a,'' and TSTF-497-A, ``Limit Inservice
Testing Program SR 3.0.2 Application to Frequencies of 2 Years or
Less.''
Date of issuance: November 5, 2010.
Effective Date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 291.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in the Federal Register: August 10, 2010 (75
FR 48375).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 5, 2010.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County,
Texas
Date of amendment request: May 27, 2010, as supplemented by letter
dated August 26, 2010.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting
Air,'' by relocating the current stored diesel fuel oil and lube oil
numerical volume and level requirements from the TSs to the TS Bases so
that it may be modified under licensee control. The TSs have been
modified so that the stored diesel fuel oil and lube oil inventory will
require that a 7-day supply be available for each diesel generator.
Condition A and Condition B in the Action table and Surveillance
Requirements (SRs) 3.8.3.1 and 3.8.3.2 are also revised to reflect the
above change. The changes are consistent with NRC-approved Revision 1
to Technical Specification Task Force (TSTF) Improved Standard
Technical Specification Change Traveler TSTF-501, ``Relocate Stored
Fuel Oil and Lube Oil Volume Values to Licensee Control.'' The
availability of the TS improvement was announced in the Federal
Register on May 26, 2010, as part of the consolidated line item
improvement process.
Date of issuance: November 4, 2010.
Effective Date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1--153; Unit 2--153.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in the Federal Register: August 10, 2010 (75
FR 48376). The supplemental letter dated August 26, 2010, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 4, 2010.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: December 17, 2009.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) for Limiting Condition for Operations
3.1.2 ``Reactivity Anomalies'' changing Surveillance Requirement
3.1.2.1 methodology.
Date of issuance: November 4, 2010.
Effective Date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 263 and 207.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the TSs.
Date of initial notice in Federal Register: February 23, 2010.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 4, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 19th day of November 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-29941 Filed 11-29-10; 8:45 am]
BILLING CODE 7590-01-P