[Federal Register Volume 76, Number 7 (Tuesday, January 11, 2011)]
[Notices]
[Pages 1644-1653]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-218]
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NUCLEAR REGULATORY COMMISSION
[NRC-2011-0005]
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 December 16 to December 29, 2010. The last
biweekly notice was published on December 28, 2010 (75 FR 81667).
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.
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. Publicly
available records will be accessible from
[[Page 1645]]
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 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 EIE, 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 1646]]
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, Room O1-F21, 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].
Calvert Cliffs Nuclear Power Plant, LLC, Docket No. 50-318, Calvert
Cliffs Nuclear Power Plant, Unit 2, Calvert County, Maryland
Date of amendment request: October 4, 2010.
Description of amendment request: The proposed amendment revises
Calvert Cliffs Technical Specification 5.5.16, ``Containment Leakage
Rate Testing Program'' to allow a one-time extension of the Type A
Integrated Leakage Rate test interval for no more than 5 years.
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?
No.
This proposed one-time extension of the Type A test interval
from 10 years to 15 years does not increase the probability of an
accident since there are no design or operating changes involved and
the test is not an accident initiator. The proposed extension of the
test interval does not involve a significant increase in the
consequences of an accident since research documented in NUREG-1493
has found that, generically, fewer than 3% of the potential
containment leak paths are not identified by Types B and C testing.
Calvert Cliffs, through testing and containment inspections, also
provides a high degree of assurance that the Containment will not
degrade in a manner detectable only by a Type A test. Inspections
required by the American Society of Mechanical Engineers Boiler and
Pressure Vessel Code are performed to identify containment
degradation that could affect leak tightness.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No.
This proposed one-time extension of the Type A test interval
from 10 years to 15 years does not involve any design or operational
changes that could lead to a new or different kind of accident from
any accident previously evaluated. The test itself is not changing
and will be performed after a longer interval. The proposed change
does not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No.
The proposed one-time extension of the Type A test interval from
10 years to 15 years does not involve a significant reduction in the
margin of safety of the containment's ability to maintain its
integrity during a design basis accident. The generic study of the
increase in the Type A test interval, NUREG-1493, concluded there is
an imperceptible increase in the plant risk associated with
extending the test interval out to 20 years. Further, the extended
test interval would have a minimal effect on this risk since Types B
and C testing detect 97% of potential leakage paths. For the
requested change in the Calvert Cliffs Integrated Leakage Rate Test
interval, it was determined that the risk contribution of leakage
will increase 0.07% (based on change in offsite dose). This change
is considered very small and does not represent a significant
reduction in the margin of safety.
Therefore, 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation,
[[Page 1647]]
Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor,
Baltimore, MD 21202.
NRC Branch Chief: Nancy L. Salgado.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: November 8, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) to eliminate provisions allowing
the High Pressure Coolant Injection (HPCI) system and the Reactor Core
Isolation Cooling (RCIC) system to be aligned to the suppression pool
when required instrument channels are inoperable. In this
configuration, the HPCI and RICI systems would not be capable of
mitigating some plant events. Also, an administrative change to the TS
Table of Contents is proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not significantly increase the
probability of an accident since it does not involve a change to any
plant equipment that initiates a plant accident. The proposed
amendment is more restrictive than the current TS in that it no
longer allows the HPCI and RCIC systems to be aligned to the
suppression pool when required instrument channels are inoperable.
The change requires HPCI and RCIC to be declared inoperable within
one hour when the associated trip functions are not operable. The
change also updates the TS Table of Contents. The HPCI system is
credited to mitigate small break loss-of-coolant accidents and the
RCIC System is not credited for accident mitigation. The proposed
change ensures the systems are aligned consistent with station
analysis assumptions. Therefore, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve any physical alteration of
plant equipment and does not change the method by which any safety-
related system performs its function. The proposed amendment is more
restrictive than the current technical specifications in that it no
longer allows the HPCI and RCIC systems to be aligned to the
suppression pool when required instrument channels are inoperable.
The change requires HPCI and RCIC to be declared inoperable within
one hour when the associated trip functions are not operable. The
change also updates the TS Table of Contents. No new or different
types of equipment will be installed and the basic operation of
installed equipment is unchanged. The methods governing plant
operation and testing remain consistent with current safety analysis
assumptions. 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.
The proposed amendment is more restrictive than the current
technical specifications in that it no longer allows the HPCI and
RCIC systems to be aligned to the suppression pool when required
instrument channels are inoperable. This ensures that safety margins
established in station safety analysis are maintained. The proposed
amendment does not involve a physical modification of the plant and
does not change the design or function of any component or system.
The proposed amendment is more restrictive than the current TS in
that it no longer allows the HPCI and RCIC systems to be aligned to
the suppression pool when required instrument channels are
inoperable. The change requires the HPCI and RCIC systems to be
declared inoperable within one hour when the associated trip
functions are not operable. The change also updates the TS Table of
Contents. This ensures analyzed safety margins are maintained.
Therefore, operation of VY in accordance with the proposed amendment
will not involve a significant reduction in the margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy Salgado.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: June 25, 2010.
Description of amendment request: The amendment would revise the
Oyster Creek Nuclear Generating Station Technical Specifications (TSs)
governing actions to be taken if a single emergency diesel generator
(EDG) is inoperable. Specifically, the proposed amendment would remove
the requirement to test the other EDG daily. Instead, the licensee
would be required to either test the other EDG once or determine that
it is not inoperable due to a common cause failure.
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 changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.]
The proposed changes are associated with the testing
requirements of the two Emergency Diesel Generators (EDGs). The
changes will eliminate unnecessary EDG testing requirements that
contribute to potential mechanical degradation of the EDGs. The
changes are based on the NRC guidance and recommendations provided
in Generic Letter (GL) 93-05, ``Line-Item Technical Specifications
Improvement to Reduce Surveillance Requirements for Testing During
Power Operation,'' and GL 94-01, ``Removal of Accelerated Testing
and Special Reporting Requirements for Emergency Diesel
Generators,'' and are consistent with NUREG-1433, ``Standard
Technical Specifications, General Electric Plants, BWR/4.'' These
proposed changes implement a recommendation promulgated in NUREG-
1366, ``Improvements To Technical Specifications Surveillance
Requirements'' to curtail daily testing of remaining operable diesel
generator[s] when one of the required diesel generators is
inoperable except for when a valid concern (e.g., potential for
common cause failure) is posed.
The probability of an accident is not increased by these changes
because the EDGs are not initiators of any design basis event.
Additionally, the proposed changes do not involve any physical
changes to plant systems, structures, or components (SSC[s]), or the
manner in which these SSC[s] are maintained [ ]. The surveillance
testing required for the limiting condition for operation for one
EDG inoperable will be eliminated for the operable EDG when the
inoperability is not due to a common cause failure. The EDG
reliability will thereby be potentially increased by reducing the
stresses on the EDG caused by unnecessary testing while maintaining
the requirement to perform a single test if a common cause failure
potentially exists. The consequences of an accident will not be
increased because the proposed changes to the EDG surveillance
requirements will continue to provide a high degree of assurance
that their operability is maintained.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. [The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.]
[[Page 1648]]
The proposed changes do not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Accordingly, the proposed changes do not introduce any
new accident initiators, nor do they reduce or adversely affect the
capabilities of any plant structure or system 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 previously evaluated.
3. [The proposed changes do not involve a significant reduction
in the margin of safety.]
The proposed changes modify the EDG accelerated testing
requirements, are consistent with NRC guidance, and [potentially]
improve EDG reliability. There are no changes being made to the
current periodic surveillance requirements. 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.
Testing the operable EDG every day for the duration of the
inoperable EDG inspection (i.e., 7 days) may be too excessive and
may lead to degradation of the EDG and possibly result in [the]
potential for unnecessary shutdowns. By reducing the possibility of
degradation from this excessive testing, the margin of safety is
[not significantly affected.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, and with the changes noted above in square brackets, 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. J. Bradley Fewell, Associate General
Counsel, Exelon Generation Company LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold Chernoff.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2 (BVPS-2), Beaver County,
Pennsylvania
Date of amendment request: February 26, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) by expanding the scope of the
steam generator (SG) tubesheet inspections using the F* inspection
methodology to the SG cold-leg tubesheet region for BVPS-2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change modifies the BVPS-2 Technical
Specifications to expand the scope of steam generator [SG] tubesheet
inspections using the F* inspection methodology to the SG cold-leg
tubesheet region based on WCAP-16385-P, Revision 1. Of the various
accidents previously evaluated in the BVPS-2 Updated Final Safety
Analysis Report (UFSAR), the proposed change only affects the SG
tube rupture (SGTR) event evaluation and the postulated steam line
break (SLB) accident evaluation. Loss-of-coolant accident (LOCA)
conditions cause a compressive axial load to act on the tube.
Therefore, since the LOCA tends to force the tube into the tubesheet
rather than pull it out, it is not a factor in this amendment
request. Another faulted load consideration is a safe shutdown
earthquake (SSE); however, the seismic analysis of Model 51M SGs has
shown that axial loading of the tubes is negligible during an SSE.
For the SGTR event, the required structural margins of the steam
generator tubes will be maintained by the presence of the tubesheet.
Tube rupture is precluded for cracks in the tube expansion region
due to the constraint provided by the tubesheet. Therefore,
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR
[pressurized-water reactor] Steam Generator Tubes,'' margins against
burst are maintained for both normal and postulated accident
conditions.
The F* length supplies the necessary resistive force to preclude
pullout loads under both normal operating and accident conditions.
The contact pressure results from the tube expansion process used
during manufacturing and from the differential pressure between the
primary and secondary side. The proposed changes do not affect other
systems, structures, components or operational features. Therefore,
the proposed change results in no significant increase in the
probability of the occurrence of an SGTR or SLB accident.
The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary
leakage flow through a postulated broken tube is not affected by the
proposed change since the tubesheet enhances the tube integrity in
the region of the expansion by precluding tube deformation beyond
its initial expanded outside diameter. The resistance to both tube
rupture and collapse is strengthened by the tubesheet in that
region. At normal operating pressures, leakage from primary water
stress corrosion cracking (PWSCC) below the F* distance is limited
by both the tube-to-tubesheet crevice and the limited crack opening
permitted by the tubesheet constraint. Consequently, negligible
normal operating leakage is expected from cracks within the
tubesheet region.
SLB leakage is limited by leakage flow restrictions resulting
from the crack and tube-to-tubesheet contact pressures that provide
a restricted leakage path above the indications and also limit the
degree of crack face opening compared to free span indications. The
total leakage (i.e., the combined leakage for all such tubes) meets
the industry performance criterion, plus the combined leakage
developed by any other alternate repair criteria, and will be
maintained below the maximum allowable SLB leak rate limit, such
that off-site doses are maintained less than 10 CFR [Title 10 of the
Code of Federal Regulation] [Part] 100 guideline values and the
limits evaluated in the BVPS-2 UFSAR.
Therefore, based on the above evaluation, the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed changes do not introduce any changes or
mechanisms that create the possibility of a new or different kind of
accident. Tube bundle integrity will continue to be maintained for
all plant conditions upon implementation of the F* methodology to
the cold-leg tubesheet region.
The proposed changes do not introduce any new equipment or any
change to existing equipment. No new effects on existing equipment
are created nor are any new malfunctions introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed changes maintain the required structural
margins of the SG tubes for both normal and accident conditions. NRC
Regulatory Guide (RG) 1.121 is used as the basis in the development
of the F* methodology for determining that SG tube integrity
considerations are maintained within acceptable limits. Regulatory
Guide 1.121 describes a method acceptable to the NRC staff for
meeting General Design Criteria 14, 15, 31, and 32. Regulatory Guide
1.121 describes the limiting safe conditions of tube wall
degradation beyond which tubes with unacceptable cracking, as
established by inservice inspection, should be removed from service
or repaired. This RG uses safety factors on loads for tube burst
that are consistent with the requirements of Section III of the
American Society of Mechanical Engineers (ASME) Code.
For primarily axially oriented cracking located within the
tubesheet, tube burst is precluded due to the presence of the
tubesheet. WCAP-16385-P, Revision 1, defines a length, F*, of
degradation-free expanded tubing that provides the necessary
resistance to tube pullout due to the pressure-induced forces (with
applicable safety factors applied). Expansion of the application of
the F* criteria to the cold-leg tubesheet region will preclude
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage provides for
large margins between calculated and actual leakage values in the F*
criteria.
Plugging of the steam generator tubes reduces the reactor
coolant flow margin for core cooling. Expansion of the F*
methodology to the cold-leg tubesheet region at BVPS-2 will result
in maintaining the
[[Page 1649]]
margin of flow that may have otherwise been reduced by tube
plugging.
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: David W. Jenkins, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: Nancy L. Salgado.
FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit No. 1 (PNPP), Lake County, Ohio
Date of amendment request: October 21, 2010.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 2.1.1, ``Reactor Core SLs,'' by
incorporating revised safety limit minimum critical power ratio
(SLMCPR) values resulting from a plant-specific analysis performed for
PNPP Cycle 14 core.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed SLMCPR value will continue to ensure that during
normal operation and abnormal operational transients, at 99.9
percent of all fuel rods in the core do not experience transition
boiling if the limit is not violated, thereby preserving the fuel
cladding integrity. The proposed TS changes do not involve any
modifications or operational changes to system, structures, or
components (SSC). The proposed TS changes do not affect any
postulated accident precursors, do not affect any accident
mitigating systems, and do no introduce any new accident initiation
mechanisms. Therefore, the proposed TS changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed TS changes do not involve any new modes of
operation, any changes to setpoints, or any plant modifications. The
proposed SLMCPR values do not result in the creation of any new
precursors to an accident. Therefore, the proposed TS changes do not
create the possibility of an accident of a different kind than
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed SLMCPR value will continue to ensure that during
normal operation and abnormal operational transients, at 99.9
percent of all fuel rods in the core do not experience transition
boiling if the limit is not violated, thereby preserving the fuel
cladding integrity. The proposed TS changes do involve modifications
or operational changes that could adversely affect the function or
performance of a SSC. The proposed TS changes do not affect any
postulated accident precursors, do not affect any accident
mitigating systems, and do not introduce any new accident initiation
mechanisms. Therefore, the proposed TS changes do not involve a
significant reduction in margin of safety.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Robert D. Carlson.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: August 5, 2010.
Description of amendment request: The proposed amendment would
modify the Callaway Plant, Unit 1, Technical Specifications (TS) by
relocating specific surveillance frequencies to a licensee-controlled
program with the guidance of Nuclear Energy Institute (NEI) 04-10,
``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed
Method for Control of Surveillance Frequencies.''
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 any accident previously
evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program [(SFCP)]. Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the technical specifications for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the surveillance requirements, and be
capable of performing any mitigation function assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly increased.
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.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
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 change involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the Final Safety Analysis Report and Bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, [the
licensee] will perform a probabilistic risk evaluation using the
guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance
with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase
of proposed changes to surveillance frequencies consistent with
Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 1650]]
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power
Station (Zion), Units 1 and 2, Lake County, Illinois
Date of amendment request: November 15, 2010.
Description of amendment request: The proposed amendments would
delete license conditions that impose specific requirements for the
decommissioning trust agreement. In lieu of the license conditions,
ZionSolutions will directly implement the requirements of 10 CFR
50.75(h)(1) through (h)(3). ZionSolutions will provide a revised trust
agreement as required by 10 CFR 50.75(h)(1)(iii) within 60 days of NRC
approval of this proposal. The licensee has stated that the trust
agreement will conform with 10 CFR 50.75(h) and ZionSolutions will take
no action under the existing trust agreement in the interim that would
be inconsistent with the provisions of the regulation.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendments alter the requirements for the
decommissioning trust fund. These revisions of the financial
assurance requirements do not involve any changes to any structures,
systems or components (SSCs) or any method of operation, maintenance
or testing. The proposed amendments will continue to provide
assurance that adequate decommissioning funding is maintained.
Changes to the terms of the trust fund will not alter previously
evaluated Defueled Safety Analysis Report (DSAR) design basis
accident assumptions, add any accident initiators, or affect the
function of the plant SSCs as to how they are operated, maintained,
modified, tested, or inspected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
Implementation of the proposed changes to decommissioning trust
fund requirements will have no impact upon the design function of
any SSC. Modifying the precise language of the administrative
controls on the fund in the trust agreement does not result in the
need for any new or different DSAR design basis accident analyses.
It does not introduce new equipment that could create a new or
different kind of accident, and no new equipment failure modes are
created. As a result, no new accident scenarios, failure mechanisms,
or limiting single failures are introduced as a result of the
proposed amendments.
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 change involve a significant reduction in a margin
of safety?
Response: No.
The margin of safety is associated with the confidence in the
ability of the fission product barriers to limit the level of
radiation to the public. The proposed amendments would not alter any
SSC functions and would not alter the way the plant is operated. The
amendments do not alter the way in which financial assurance for
decommissioning is achieved. The proposed amendments would not
introduce any new uncertainties associated with any safety limit.
The proposed amendments would have no impact upon the structural
integrity of the fuel cladding or any other barrier to fission
product release. There would be no reduction in the effectiveness of
the fission product barriers to limit the level of radiation to the
public. Therefore, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Russ Workman, Deputy General Counsel,
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT
84101.
NRC Branch Chief: Bruce Watson.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management 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 8, 2010.
Brief description of amendment: The amendments deleted redundant
reporting and operational restriction provisions from Technical
Specification (TS) Section 2.2, ``Safety Limit Violations,'' consistent
with Technical Specification Task Force (TSTF) change traveler TSTF-5-
A, Revision 1, ``Delete Safety Limit Violation Notification
Requirements,'' and replaced plant-specific titles with generic titles
in TS Section 5.2.1, ``Onsite and Offsite Organizations,'' consistent
with TSTF-65-A, Revision 1, ``Use of Generic Titles for Utility
Positions.''
Date of issuance: December 29, 2010.
[[Page 1651]]
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 1--183; Unit 2--183; Unit 3--183.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: July 27, 2010 (75 FR
44022).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 29, 2010.
No significant hazards consideration comments received: No.
Carolina Power and Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: July 21, 2009, as supplemented
March 3 and July 28, 2010.
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 6.9.1.6 to add NRC approved Topical Report
(TR) EMF-2310(P)(A), ``SRP Chapter 15 Non-LOCA Methodology for
Pressurized Water Reactors,'' to the Core Operating Limits Report
methodologies list. This change will allow the use of thermal-hydraulic
analysis code S-RELAP5 for Final Safety Analysis Report (FSAR) Chapter
15 non-loss-of-coolant accident (LOCA) transients in the HNP safety
analyses. TR EMF-2310(P)(A), Revision 0, was approved by the NRC on May
11, 2001, for the application of the S-RELAP5 thermal-hydraulic
analysis computer code to FSAR Chapter 15 non-LOCA transients. EMF-
2310(P)(A), Revision 1, approved by the NRC on May 19, 2004, updated
Section 5.6 of the TR.
Date of issuance: December 23, 2010.
Effective date: Effective as of the date of issuance and shall be
implemented within 60 days.
Amendment No.: 135.
Renewed Facility Operating License No. NPF-63: The amendment
revises the TSs and facility operating license.
Date of initial notice in Federal Register: November 10, 2009 (74
FR 58060). The supplements dated March 3, and July 28, 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 amendment is contained
in a safety evaluation dated December 23, 2010.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: December 14, 2009, as
supplemented by letters dated September 8, 2010, and October 28, 2010.
Brief description of amendments: The amendments revised the
Technical Specifications by revising Surveillance Requirements 3.8.4.3
and 3.8.4.6. These TS SRs address battery connection resistance values.
Date of issuance: December 20, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 262, 258.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: August 10, 2010 (75 FR
48375). The supplements dated September 8, 2010, and October 28, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 20, 2010.
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: December 14, 2009, as
supplemented by letters dated September 8, 2010, and October 28, 2010.
Brief description of amendments: The amendments revised the
Technical Specifications by revising Surveillance Requirements 3.8.4.2
and 3.8.4.5. These TS SRs address battery connection resistance values.
Date of issuance: December 20, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 260, 240.
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: August 10, 2010 (75 FR
48375). The supplements dated September 8, 2010, and October 28, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 20, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2 (Braidwood), Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2
(Byron), Ogle County, Illinois
Date of application for amendment: December 16, 2009, as
supplemented by letters dated April 26 and October 25, 2010.
Brief description of amendment: The amendments revise Technical
Specifications Section 5.6.5, ``Core Operating Limits Report,'' to
replace the existing reference for the large break loss-of-coolant
accident (LOCA) analysis methodology with a reference to WCAP-16009-P-
A, Revision 0, ``Realistic Large Break LOCA Evaluation Methodology
Using the Automated Statistical Treatment of Uncertainty Method,''
January 2005.
Date of issuance: December 21, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: Braidwood Unit 1--164; Braidwood Unit 2--164; Byron
Unit No. 1--170; and Byron Unit No. 2--170.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66:
The amendments revise the TSs and Licenses.
Date of initial notice in Federal Register: February 23, 2010 (75
FR 8141). The supplemental letters dated April 26, and October 25,
2010, contained clarifying information, did not change the initial no
significant hazards consideration determination, and did not expand the
scope of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 21, 2010.
[[Page 1652]]
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit 1 and 2, St.. Lucie County, Florida.
Date of application for amendments: December 14, 2009, as
supplemented on July 30, 2010.
Brief description of amendments: Amendment modifies Technical
Specification (TS) 3/4 .4.10 ``Structural Integrity,'' in Unit 1 (TS 3/
4.4.11 in Unit 2), TS 3.3.3.8, ``Accident Monitoring Instrumentation,''
in Unit 1 (TS 3.3.3.6 in Unit 2), TS 6.4.1, ``Training,'' in Units 1
and 2, and several administrative changes in the TSs for both units .
The changes delete the Structural Integrity TS, update Accident
Monitoring Instrumentation requirements and make various administrative
TS changes.
Date of Issuance: December 28, 2010.
Effective Date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 210, 159.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the TSs.
Date of initial notice in Federal Register: April 20, 2010 (75 FR
20638). The supplement dated July 30, 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 December 28, 2010.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-354, 50-272 and 50-311, Hope Creek
Generating Station and Salem Nuclear Generating Station, Unit 1 and 2,
Salem County, New Jersey
Date of application for amendments: March 25, 2010.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) associated with reactor coolant system
(RCS) structural integrity requirements for Hope Creek Generating
Station (HCGS) and Salem Nuclear Generating Station (Salem), Unit Nos.
1 and 2. Specifically, the amendments revise the TSs to: (1) Delete the
RCS structural integrity requirements contained in HCGS TS 3/4.4.8,
Salem Unit 1 TS 3/4.4.10, and Salem Unit 2 TS 3/4.4.11; (2) relocate
the augmented inservice inspection requirements for the reactor coolant
pump flywheel, currently contained in Salem Unit 1 surveillance
requirement (SR) 4.4.10.1.1 and Salem Unit 2 SR 4.4.11.1, to a new
program in TS 6.8.4.k; and (3) delete the augmented inservice
inspection program requirements for the steam generator channel heads
currently contained in Salem Unit 1 SR 4.4.10.1.2.
Date of issuance: December 15, 2010.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 186, 298 and 281.
Facility Operating License Nos. NPF-57, DPR-70 and DPR-75: The
amendments revised the TSs and the Licenses.
Date of initial notice in Federal Register: June 15, 2010 (75 FR
33843).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 15, 2010.
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 amendment: January 26, 2010 (TS 09-05).
Brief description of amendment: The amendments revised the
Technical Specification (TS) Table 3.3-1, ``Reactor Trip System
Instrumentation,'' Functional Unit 5, ``Intermediate Range, Neutron
Flux,'' to resolve an oversight regarding the operability requirements
for the intermediate range neutron flux channels. The amendments added
an action to TS Table 3.3-1 to define that the provisions of
Specification 3.0.3 are not applicable above 10 percent of thermal
rated power with the number of operable intermediate range neutron flux
channels two less than the minimum channels operable requirement.
Date of issuance: December 21, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 328, 321.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: March 23, 2010 (75 FR
13791).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated December 21, 2010.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: February 10, 2010.
Brief Description of amendments: These amendments revise the
Technical Specifications 5.2.1, ``Fuel Assemblies,'' to add Optimized
ZIRLO\TM\ as an acceptable fuel rod cladding material. In addition, the
amendments propose adding the Westinghouse topical report for Optimized
ZIRLO\TM\ to the analytical methods used to determine the core
operating limits listed in TS 6.2.C.
Date of issuance: December 22, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 271, 270.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the licenses and the technical specifications.
Date of initial notice in Federal Register: August 27, 2010 (75 FR
52781).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 22, 2010.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 4, 2009, as supplemented by
letters dated March 25 and November 17, 2010.
Brief description of amendment: The amendment revised the approved
fire protection program as described in the Wolf Creek Generating
Station (WCGS) Updated Safety Analysis Report (USAR). Specifically, a
deviation from certain technical requirements of Title 10 of the Code
of Federal Regulations (10 CFR), part 50, appendix R, section III.G.2,
as documented in Appendix 9.5E of the WCGS USAR, was requested
regarding the use of operator manual actions in lieu of meeting circuit
separation protection criteria. Table 3-1 of the submittal dated March
4, 2009 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML090771269), identified the proposed feasible and
reliable operator manual actions requested for permanent approval and
Table 3-2 of the submittal identified the proposed feasible operator
manual actions requested for approval on an interim basis. The interim
operator actions will be eliminated with the
[[Page 1653]]
implementation of associated design change package. The amendment also
revised license condition 2.C.(5)(a) to include the deviation approved
by the amendment request.
Date of issuance: December 16, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 191.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 21, 2009 (75 FR
18258). The supplemental letters dated March 25 and November 17, 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 amendment is contained
in a Safety Evaluation dated December 16, 2010.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 16, 2009, as supplemented by
letter dated August 26, 2010.
Brief description of amendment: The amendment revised the battery
acceptance criteria in Technical Specification 3.8.4, ``DC [Direct
Current] Sources--Operating,'' Surveillance Requirements (SRs) 3.8.4.2
and 3.8.4.5. Specifically, the amendment modified SR 3.8.4.2 and SR
3.8.4.5 by providing limits for inter-cell, inter-tier/inter-bank/
terminal, and field jumper connections for 60-cell, 59-cell, and 58-
cell configurations.
Date of issuance: December 20, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 192.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 6, 2010 (75 FR
17448). 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 amendment is contained
in a Safety Evaluation dated December 20, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 30th day of December 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2011-218 Filed 1-10-11; 8:45 am]
BILLING CODE 7590-01-P