[Federal Register Volume 75, Number 248 (Tuesday, December 28, 2010)]
[Notices]
[Pages 81667-81675]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-32668]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0393]
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 2, 2010, to December 15, 2010. The
last biweekly notice was published on December 14, 2010 (75 FR 77906).
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, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the 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
[[Page 81668]]
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 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 (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
[[Page 81669]]
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 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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: July 22, 2010.
Description of amendment request: The amendments would revise an
element of the methodology used in evaluating the radiological
consequences of design basis steam generator tube rupture (SGTR)
accidents. Specifically, the changes will revise the Palo Verde Nuclear
Generating Station (PVNGS) Updated Final Safety Analysis Report
(UFSAR), Section 15.6.6, ``Steam Generator Tube Rupture,'' to reflect a
lower iodine spiking factor assumed for the coincident event Generated
Iodine Spike (GIS) and the resulting reduction in the radiological
consequences provided in UFSAR Table 15.6.3-5, ``Radiological
Consequences for the Limiting SGTRLOPSF [Steam Generator Tube Rupture
with Loss of Offsite Power and Single Failure] Event.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment changes an element of the methodology
used in evaluating the radiological consequences of design basis
SGTR accidents. This change will revise the iodine spiking factor
used for a GIS from a value of 500 to a value of 335. The proposed
change in the methodology element does not involve any design or
physical changes to the facility or any component of that facility.
The proposed change creates no new failure modes or initiating
occurrences that could result in a design basis transient or
accident evaluated in the Palo Verde Nuclear Generating Station
(PVNGS) Updated Final Safety Analysis Report (UFSAR). Therefore the
proposed change does not involve a significant increase in the
probability of an accident previously evaluated.
The proposed change in the methodology element does change the
design basis analyses results for PVNGS. However, the results remain
bounded by the previous analyzed values and remain within the
acceptance criteria for PVNGS of 100% of the 10 CFR [Part] 100
maximum thyroid dose limit of 300 rem [roentgen equivalent man].
Therefore, the proposed change does not involve a significant
increase in the consequences of an accident previously analyzed.
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 amendment changes an element of the methodology
used in evaluating the radiological consequences of design basis
SGTR accidents. This change will revise the iodine spiking factor
used for a GIS from a value of 500 to a value of 335. The proposed
change in the methodology element does not involve any design or
physical changes to the facility or any component of that facility.
The proposed change in the methodology element does change the
design basis analyses results for PVNGS; however, these results
remain bounded by the previous analyzed values and remain within the
acceptance criteria for PVNGS of 100% of the 10 CFR [Part] 100
maximum thyroid dose limit of 300 rem.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment changes an element of the methodology
used in evaluating the radiological consequences of design basis
SGTR accidents. This change will revise the iodine spiking factor
used for a GIS from a value of 500 to a value of 335. The proposed
change in the methodology element does not involve any design or
physical changes to the facility or any component of that facility.
The proposed methodology element change for a postulated SGTR, with
a coincident loss of offsite power, GIS, and a failed open
atmospheric dump valve (ADV), results in lower maximum dose
consequences at the Exclusion Area Boundary (EAB) and Low Population
Zone (LPZ) [than] previously analyzed for this event combination.
The methodology element change results in the 2-hour maximum thyroid
dose value of 182 rem at the EAB being reduced to 124 rem. In
addition, the 8-hour maximum thyroid dose of 125 rem at the LPZ,
would be reduced to 84 rem.
Previously for PVNGS, the GIS 8-hour maximum thyroid dose was
bounding at the LPZ and the pre-Accident Iodine Spike (PIS) 2-hour
maximum thyroid dose was bounding at the EAB. The methodology
element change reduces the GIS calculated dose at both the EAB and
LPZ for SGTR events, but it does not affect the PIS dose values.
Since the GIS calculated dose at the LPZ drops below the PIS 8-hour
LPZ maximum thyroid dose (91 rem), the PIS 8-hour LPZ dose will
become bounding for PVNGS. The PIS 2-hour EAB maximum thyroid dose
(294 rem), remains the bounding dose at the EAB.
The revised dose consequences remain bounded by the previous
analyzed values and remain within the 10 CFR Part 100 guideline
values which are the acceptance criteria for PVNGS Units 1, 2, and
3. In addition, the proposed change has no effect on previously
reported dose consequences for control room personnel following any
postulated SGTR event.
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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: October 6, 2010.
Description of amendment request: The proposed change will revise
the note in Surveillance Requirement (SR) 3.5.4.1 in the Refueling
Water Storage
[[Page 81670]]
Tank (RWST) Technical Specification (TS). Specifically, the proposed
change will not require monitoring of the RWST temperature every 24
hours when the RWST heating steam supply isolation valves are locked
closed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change revises the existing Indian Point
[Nuclear Generating Unit No.] 3 [(IP3)] Refueling Water Storage Tank
(RWST) Technical Specification (TS) Surveillance Requirement (SR)
3.5.4.1 to revise the note that eliminates the requirement to
perform SR 3.5.4.1 when ambient air temperatures are within the
operating limits of the RWST. The revision to the note adds a
requirement that the steam heating supply isolation valves be locked
closed when not performing the surveillance. The additional
requirement does not increase the probability of an accident
occurring since it is not an accident initiator and does not
increase the consequences of an accident since it is providing
additional assurance that the RWST is within the temperature limits
assumed for accident analyses. The change increases observation of
the RWST temperature when the steam supply isolation valves are not
locked closed and does not otherwise affect [* * *] the performance
capability of the structures, systems, and components relied upon to
mitigate the consequences of postulated accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change revises the note that eliminates the
requirement to perform SR 3.5.4.1 when ambient air temperatures are
within the operating limits of the RWST. The revision adds the
additional requirement of locking closed the steam supply isolation
valves. The proposed change does not involve installation of new
equipment or modification of existing equipment, so that no new
equipment failure modes are introduced. Also, the proposed change
does not result in a change to the way that the equipment or
facility is operated so that no new accident initiators are created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change revises the note that eliminates the
requirement to perform SR 3.5.4.1 when ambient air temperatures are
within the operating limits of the RWST. The revision adds the
additional requirement of locking closed the steam supply isolation
valves. The change does not reduce margin since it increases the
temperature surveillance frequency for the RWST to provide further
assurance that the required water temperature is maintained at all
times.
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: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy L. Salgado.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: November 8, 2010.
Description of amendment request: The proposed license amendment
request will make changes related to the final resolution of an
unresolved issue associated with Technical Specification (TS) Amendment
No. 181 dated February 25, 2009. This issue was resolved with the
approval of Revision 4 of Technical Specification Task Force (TSTF)
Change Traveler TSTF-493, ``Clarify Application of Setpoint Methodology
for LSSS [Limiting Safety System Setting] Functions,'' which included
the instrument function (i.e., Condensate Storage Tank (CST) Level-Low)
that was the subject of Amendment No. 181. Specifically, the proposed
change will add the appropriate notes as specified in TSTF-493 to the
surveillance requirements associated with TS Table 3.3.5.1-1,
``Emergency Core Cooling System Instrumentation,'' Function 3.d,
Condensate Storage Tank Level--Low, and to TS Table 3.3.5.2-1,
``Reactor Core Isolation Cooling System Instrumentation,'' Function 3,
Condensate Storage Tank Level--Low. The supporting TS Bases will also
be revised.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change adds test requirements to the CST Level-Low
function to ensure the CST Level-low instruments will function as
required. Surveillance tests are not an initiator of any accident
previously evaluated. The CST components, for which the additional
requirements were added, continue to be operable and capable of
performing their intended 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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical change to the
plant, i.e., no new or different type of equipment will be
installed. The proposed change does not alter assumptions made in
the safety analysis but ensures that the CST Level-low instruments
perform as assumed in the [Updated Final Safety Analysis Report].
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adds test requirements that will assure that
(1) the CST Level-low instrumentation for the setpoint allowable
value will be the limiting setting for assessing instrumentation
channel operability and (2) will be conservatively determined so
that the evaluation of CST instrument performance history and the
requirements of the calibration procedures will not have an adverse
effect on equipment operability. The testing methods and acceptance
criteria for the CST Level-low instrumentation will continue to be
met. There is no impact to the safety analysis acceptance criteria
as described in the plant licensing basis because no change is made
to the accident analysis assumptions.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
[[Page 81671]]
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 12, 2010.
Description of amendment request: The proposed amendment would
modify Item 1 of Table 2-5, ``Instrumentation Operating Requirements
for Other Safety Feature Functions,'' of Technical Specification (TS)
2.15, ``Instrumentation and Control Systems,'' to provide new Note (e),
and Surveillance Requirement (SR) Items 1 and 2 of Table 3-3, ``Minimum
Frequencies for Checks, Calibrations and Testing of Miscellaneous
Instrumentation and Controls,'' of TS 3.1, ``Instrumentation and
Control,'' which pertain to operability of the primary and secondary
control element assembly (CEA) position indication system (CEAPIS)
channels. A new SR is proposed for Item 4 of Table 3-3 of TS 3.1, which
will verify the position of CEAs each shift. The proposed amendment
will ensure that CEA alignment is maintained during power operations so
that the power distribution and reactivity limits defined by the design
power peaking and shutdown margin (SDM) limits are preserved.
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 will allow plant operation to continue
when a CEAPIS channel is inoperable by requiring prompt verification
of CEA positions following CEA movement. CEAs are most likely to
become misaligned during movement and therefore, this change will
cause CEA alignment errors to be promptly detected and corrected. It
is appropriate to clarify that CEAPIS channels are not subject to
the requirements of TS 2.15(1), (2), and (3) as they are not
designed to be placed in trip or bypass, nor are they engineered
safety feature (ESF) or isolation logic subsystems.
The proposed amendment does not alter the requirements of TS
2.15(4) regarding the rod block function of the secondary CEAPIS
channel. Should the secondary CEAPIS channel or its rod block
function be inoperable, several additional CEA deviation events are
possible. However, this situation is already addressed by TS
2.15(4), which requires the CEAs (rods) to be maintained fully
withdrawn with the control rod drive system mode switch in the off
position except when manual motion of CEA Group 4 is required to
control axial power distribution. This is the same position that the
CEAs must be in (fully withdrawn) when the plant is at power (Mode
1) in order to utilize distributed control system (DCS) core mimic
to CHANNEL CHECK the CEAPIS channels.
If it was not possible to use DCS core mimic to verify the
primary CEAPIS channel as would be the case if CEA Group 4 was
inserted to control axial power distribution, then the primary
CEAPIS channel would be declared inoperable when the CHANNEL CHECK
could not be accomplished. The plant would then be placed in hot
shutdown (Mode 3) within 12 hours in accordance with TS 2.15(4).
Therefore, although the proposed amendment will allow a CEAPIS
channel to be inoperable indefinitely, there is no significant
increase in the probability or consequences of an accident as the
requirements of TS 2.15(4) will continue to be met. This serves to
prevent the type of CEA deviation events that the rod block function
was designed for.
Replacing the current method of verifying CEAPIS data with the
defined term CHANNEL CHECK is an improvement that provides
additional flexibility without weakening the intent of the
surveillance. As a result, when it is feasible to obtain CEA
position indication from DCS core mimic (i.e., when the CEAs are
either fully inserted or fully withdrawn), the primary and secondary
CEAPIS channels will be compared with DCS core mimic indication as
well as each other.
As an additional means of verifying CEA positions, DCS core
mimic indication provides added confidence that the CEAs are in the
indicated positions. Should the primary or secondary CEAPIS channel
become inoperable, the accuracy and reliability of DCS core mimic
indication is assured by its previous comparison with both OPERABLE
channels. Comparison of the OPERABLE CEAPIS channel with DCS core
mimic will satisfy the required CHANNEL CHECK and allow continued
operation while the inoperable channel is repaired. The proposed
amendment ensures that the CEA alignment required by TS 2.10.2(4) is
met each shift by requiring all full length (shutdown and
regulating) CEAs to be positioned within 12 inches of all other CEAs
in the group.
The change proposed for TS 2.10.2(7)c incorporates more
conservative wording to ensure that the regulating CEA groups are
maintained within the Long Term Insertion Limit. The proposed change
will ensure that corrective actions are taken if either time
interval is exceeded and makes TS 2.10.2(7)c more consistent with CE
STS.
The proposed amendment does not alter the plant configuration,
require new plant equipment to be installed, alter accident analysis
assumptions, add any initiators, or affect the function of plant
systems or the manner in which systems are operated, maintained,
modified, tested, or inspected. As an additional means of verifying
primary and secondary CEAPIS data, DCS core mimic indication
increases confidence in the reliability of CEAPIS data.
The proposed amendment will help minimize unplanned shutdowns
that can cause plant transients yet continues to ensure that power
distribution and reactivity limits are maintained.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not change the design function or
operation of the primary or secondary CEAPIS channels. If one CEAPIS
channel should become inoperable, the position of CEAs will be
verified within 15 minutes of any CEA movement to quickly detect and
correct CEA alignment errors. Data from each CEAPIS channel will
continue to be compared to the other channel each shift as before.
However, a CHANNEL CHECK will require that CEAPIS channel data also
be compared with DCS core mimic indication when it is available.
Thus, when the CEAPIS channels are required to be OPERABLE, there
will be at least two means of verifying the position of CEAs or else
appropriate actions must be taken. The CEA alignment required by TS
2.10.2(4) is assured by requiring verification each shift that all
full length (shutdown and regulating) CEAs are positioned within 12
inches of all other CEAs in the group.
No changes are proposed to testing and calibration of the CEAPIS
channels and these requirements will continue to ensure that they
are capable of performing their design function. Use of the defined
term CHANNEL CHECK is an appropriate surveillance method as it
requires that the channel be compared with other independent
channels measuring the same variable where feasible. DCS core mimic
is a diverse, accurate and reliable means of verifying CEA positions
when the CEAs are fully inserted or fully withdrawn. The change
proposed for TS 2.10.2(7)c ensures that appropriate corrective
actions are taken when the regulating CEA groups are below the Long
Term Insertion Limit in excess of either of the specified time
intervals.
No new structures, systems, or components (SSCs) are being
installed, and no credible new failure mechanisms, malfunctions, or
accident initiators are created.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
When a CEAPIS channel is inoperable, the proposed amendment
allows plant operation to continue but requires more frequent
verification of CEA positions following any CEA movement, which is
when CEAs are most likely to become misaligned. This will enable CEA
alignment errors to be detected and corrected more promptly. As
CEAPIS channels are not designed to be placed in trip
[[Page 81672]]
or bypass, nor are they engineered safety feature (ESF) or isolation
logic subsystems, it is appropriate to clarify that TS 2.15(1), (2),
and (3) do not apply. FCS normally operates with the CEAs fully
withdrawn and maintains reactivity control by adjusting reactor
coolant system (RCS) boric acid concentration. When the CEAs are
fully withdrawn (or fully inserted), DCS core mimic indication
provides accurate and reliable indication of CEA positions suitable
for comparison with the primary and secondary CEAPIS channels. Thus,
even with one CEAPIS channel inoperable, a diverse means of
verifying the accuracy of the OPERABLE CEAPIS channel will be
available. The accuracy and reliability of DCS core mimic is assured
by testing conducted each refueling outage with continued assurance
provided by comparison with primary and secondary CEAPIS each shift.
The change also ensures that the CEA alignment required by TS
2.10.2(4) is met each shift by requiring all full length (shutdown
and regulating) CEAs to be positioned within 12 inches of all other
CEAs in the group. The proposed amendment does not alter the TS
2.15(4) requirement to place the reactor in hot shutdown in the
event that both CEAPIS channels are inoperable. The change proposed
for TS 2.10.2(7)c incorporates more conservative wording to ensure
that the regulating CEA groups are maintained within the Long Term
Insertion Limit.
The proposed amendment will help minimize unplanned shutdowns
that can cause plant transients yet continues to ensure that power
distribution and reactivity limits are maintained. The proposed
amendment does not alter the plant configuration, require new plant
equipment to be installed, alter accident analysis assumptions, add
any initiators, or affect the function of plant systems or the
manner in which systems are operated, maintained, modified, tested,
or inspected.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: October 4, 2010.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) requirements for snubbers in TS
3/4.7.9 due to planned revisions to the inservice inspection (ISI)
program.
For the current third 10-year ISI intervals, at Salem Nuclear
Generating Station (Salem), Units 1 and 2, snubber testing and
examination are performed in accordance with the specific requirements
of TS 3/4.7.9 in lieu of the requirements contained in American Society
of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),
Section XI, Article IWF-5000, as previously authorized by the U.S.
Nuclear Regulatory Commission (NRC or the Commission).
Section 50.55a(g)(4)(ii) of Title 10 of the Code of Federal
Regulations (10 CFR) requires that inservice examination of components
conducted during successive 120-month inspection intervals must comply
with the requirements of the latest edition and addenda of the ASME
Code incorporated by reference in 10 CFR 50.55a(b), 12 months before
the start of the inspection interval. For the Salem Unit 1 fourth 10-
year ISI interval beginning on May 20, 2011, the licensee intends to
adopt Subsection ISTD of the ASME Code for Operation and Maintenance of
Nuclear Power Plants (OM Code), 2004 Edition, in place of the
requirements for snubbers in ASME Code, Section XI, Articles IWF-
5200(a) and (b) and IWF-5300(a) and (b), as permitted by 10 CFR
50.55a(b)(3)(v). The licensee also intends to adopt Subsection ISTD of
the ASME OM Code for the remainder of the Salem Unit 2 third 10-year
ISI interval which ends on November 27, 2013.
In accordance with 10 CFR 50.55a(g)(5)(ii), if a revised ISI
program for a facility conflicts with the TSs for the facility,
licensees are required to apply to the Commission for amendment of the
TSs to conform the TSs to the revised program. Due to the planned
changes to the ISI program, the proposed amendment would replace the
specific TS requirements for snubbers, currently contained in
surveillance requirement (SR) 4.7.9, with reference to the program for
examination, testing and service life monitoring for snubbers. In
addition, the current reference to SR 4.7.9c in TS ACTION 3.7.9 would
be replaced with reference to the program for snubbers.
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 Nuclear Regulatory
Commission (NRC) staff edits in square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise [TS 3/4.7.9 due to planned changes
to the ISI program for snubbers. Specifically, the proposed
amendment would replace the TS SRs for snubbers with reference to
the program for examination, testing and service life monitoring for
snubbers. Following implementation of the proposed amendment, in
lieu of the TS SRs, snubber examination, testing and service life
monitoring would be governed by the requirements in Section XI of
the ASME Code or the OM Code as required by 10 CFR 50.55a(g) or 10
CFR .55a(b)(3)(v), except where the NRC has granted specific written
relief, pursuant to 10 CFR 50.55a(g)(6)(i), or authorized
alternatives pursuant to 10 CFR 50.55a(a)(3).]
Snubber examination, testing and service life monitoring is not
an initiator of any accident previously evaluated. Therefore, the
probability of an accident previously evaluated is not significantly
increased.
Snubbers will continue to be demonstrated OPERABLE by
performance of a program for examination, testing and service life
monitoring in compliance with 10 CFR 50.55a or authorized
alternatives. The proposed change to TS ACTION 3.7.9 for inoperable
snubbers is administrative in nature and is required for consistency
with the proposed change to SR 4.7.9. Therefore the proposed change
does not adversely affect plant operations, design functions or
analyses that verify the capability of systems, structures, and
components to perform their design functions. The consequences of
accidents 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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve any physical alteration of
plant equipment. The proposed change does not change the method by
which any safety-related system performs its function. As such, 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 previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes ensure snubber examination, testing and
service life monitoring will continue to meet the requirements of 10
CFR 50.55a(g) except where the NRC has granted specific written
relief, pursuant to 10 CFR 50.55a(g)(6)(i), or authorized
alternatives pursuant to 10 CFR
[[Page 81673]]
50.55a(a)(3). Snubbers will continue to be demonstrated OPERABLE by
performance of a program for examination, testing and service life
monitoring in compliance with 10 CFR 50.55a or authorized
alternatives. The proposed change to TS ACTION 3.7.9 for inoperable
snubbers is administrative in nature and is required for consistency
with the proposed change to SR 4.7.9.
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, 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: Vincent Zabielski, PSEG Nuclear LLC-N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 22, 2010, as supplemented by
letter dated November 22, 2010.
Description of amendment request: The proposed amendment consists
of changes to the approved fire protection program as described in the
Wolf Creek Generating Station (WCGS) Updated Safety Analysis Report
(USAR). Specifically, amendment proposes a deviation from a commitment
to certain technical requirements of 10 CFR, Part 50, Appendix R,
Section III.L.1, as described in Appendix 9.5E of the WCGS USAR. The
licensee has proposed to revise USAR Table 9.5E-1 to include
information on Reactor Coolant System process variables not maintained
within those predicted for a loss of normal ac [alternating current]
power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design function of structures, systems and components (SSCs)
are not impacted by the proposed change. Evaluation SA-08-006 Rev. 1
[RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation
for a Postulated Control Room Fire] has demonstrated that the
formation of voids in the reactor head for a short time following a
fire in the control room and spurious temporary opening of the
pressurizer power operated relief valve (PORV) does not result in
damage to a fission product barrier and does not result in a loss of
natural circulation cooldown. The proposed change does not alter or
prevent the ability of SSCs from performing their intended function
to mitigate the consequences of an initiating event within the
assumed acceptance limits.
Therefore, the probability of any accident previously evaluated
is not increased. Equipment required to mitigate an accident remains
capable of performing the assumed function.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change will not alter the requirements or function
for systems required during accident conditions. The design function
of structures, systems and components are not impacted by the
proposed change. The thermal hydraulic analysis of the reactor
coolant system identified that the process variables are not
maintained within those predicted for a loss of normal ac power,
however, the fission product boundary integrity is not affected.
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 will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on departure from
nuclear boiling ratio (DNBR) limits, heat flux hot channel factor
(FQ(Z)) limits, nuclear enthalpy rise hot channel factor
(FNN[Delta]H) limits, peak centerline
temperature (PCT) limits, peak local power density or any other
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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: November 4, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.6.5, ``CORE OPERATING LIMITS
REPORT (COLR),'' to replace the existing large break loss-of-coolant
accident (LOCA) analysis methodology. Specifically, the proposed change
adds a reference of Westinghouse Electric Company's topical report
WCAP-16009-P-A, ``Realistic Large Break LOCA Evaluation Methodology
Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),''
to TS 5.6.5b.
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 change revises TS Section 5.6.5 to incorporate a
new large break LOCA analysis methodology. Specifically, the
proposed change adds WCAP-16009-P-A to TS 5.6.5b as a method used
for establishing core operating limits.
Accident analyses are not accident initiators; therefore, the
proposed change does not involve a significant increase in the
probability of an accident. The analyses using ASTRUM demonstrated
that the acceptance criteria in 10 CFR 50.46, ``Acceptance criteria
for emergency core cooling systems for lightwater nuclear power
reactors,'' were met. Large break LOCA analyses performed consistent
with the methodology in NRC-approved WCAP-16009-P-A, including
applicable assumptions, limitations and conditions, demonstrate that
10 CFR 50.46 acceptance criteria are met; thus, this change does not
involve a significant increase in the consequences of an accident.
No physical changes to the plant are associated with the proposed
change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed change revises TS Section 5.6.5 to incorporate a
new large break LOCA analysis methodology. Specifically, the
proposed change adds WCAP-16009-P-A to TS 5.6.5b as a method used
for establishing core operating limits. There are no physical
changes being made to the plant as a result
[[Page 81674]]
of using the Westinghouse ASTRUM analysis methodology in WCAP-16009-
P-A for performance of the large break LOCA analyses. Large break
LOCA analyses performed consistent with the methodology in NRC-
approved WCAP-16009-P-A, including applicable assumptions,
limitations and conditions, demonstrate that 10 CFR 50.46 acceptance
criteria are met. No new modes of plant operation are being
introduced. The configuration, operation, and accident response of
the structures or components are unchanged by use of the new
analysis methodology. Analyses of transient events have confirmed
that no transient event results in a new sequence of events that
could lead to a new accident scenario. The parameters assumed in the
analyses are within the design limits of existing plant equipment.
In addition, employing the Westinghouse ASTRUM large break LOCA
analysis methodology does not create any new failure modes that
could lead to a different kind of accident. The design of systems
remains unchanged and no new equipment or systems have been
installed which could potentially introduce new failure modes or
accident sequences. No changes have been made to instrumentation
actuation setpoints. Adding the reference to WCAP-16009-P-A in TS
Section 5.6.5b is an administrative change that does not create the
possibility of a new or different kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises TS Section 5.6.5 to incorporate a
new large break LOCA analysis methodology. Specifically, the
proposed change adds WCAP-1 6009-P-A to TS 5.6.5b as a method used
for establishing core operating limits. The analyses using ASTRUM
demonstrated that the applicable acceptance criteria in 10 CFR 50.46
are met. Margins of safety for large break LOCAs include
quantitative limits for fuel performance established in 10 CFR
50.46. These acceptance criteria are not being changed by this
proposed new methodology. Large break LOCA analyses performed
consistent with the methodology in NRC-approved WCAP-16009-P-A,
including applicable assumptions, limitations and conditions,
demonstrate that 10 CFR 50.46 acceptance criteria are met; thus,
this change does not involve a significant 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
NextEra Energy Point Beach, Docket Nos. 50-266 and 50-301, Point Beach
Nuclear Plant (PBNP), Units 1 and 2, Manitowoc County, Wisconsin
Date of amendment request: April 7, 2009, as supplemented by
letters dated June 17 (two letters) and December 8 of 2009; and April
15, July 8, July 28, August 24, September 9, September 21, October 14,
and November 1 of 2010.
Brief description of amendment request: The proposed amendment
would increase the licensed core power level for PBNP Units 1 and 2
from 1540 to 1800 megawatts thermal. The increase in core thermal power
will be approximately 17 percent over the current licensed thermal
power level and is categorized as an Extended Power Uprate.
Date of publication of individual notice in Federal Register:
November 17, 2010 (75 FR 70305).
Expiration date of individual notice: January 18, 2011.
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 01 F21, 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].
Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendment: September 8, 2010.
Brief description of amendment: The amendments delete the Technical
Specification requirements related to the containment hydrogen
recombiners and the hydrogen monitors, in accordance with Nuclear
Energy Institute Technical Specification Task Force (TSTF) initiative
designated as TSTF-447.
Date of issuance: December 14, 2010.
Effective date: As of the date of issuance and shall be implemented
[[Page 81675]]
within 120 days from the date of issuance.
Amendment Nos.: 313 (for Unit 1) and 296 (for Unit 2).
Facility Operating License Nos. DPR-58 and DPR-74: Amendment
revised the Renewed Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 14, 2010 (75 FR
63209).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 14, 2010.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of application for amendments: November 24, 2009, as
supplemented by letter dated May 26, 2010.
Brief description of amendments: These amendments revise Technical
Specification (TS) 4.2.1, ``Fuel Assemblies,'' to add Optimized
ZIRLO\TM\ as an acceptable fuel rod cladding material and add two
Westinghouse topical reports to the analytical methods identified in TS
5.6.5.b.
Date of issuance: November 29, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 199, 187.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 4, 2010 (75 FR
23816).
The supplemental letter contained clarifying information and 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 November 29, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland this 16th 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. 2010-32668 Filed 12-27-10; 8:45 am]
BILLING CODE 7590-01-P