[Federal Register Volume 75, Number 65 (Tuesday, April 6, 2010)]
[Notices]
[Pages 17439-17452]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-7451]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0145]
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 March 11, 2010, to March 24, 2010. The last
biweekly notice was published on March 23, 2010 (75 FR 13786).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking
and Directives Branch (RDB), TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be faxed to the RDB at 301-492-3446. Documents may be examined, and/or
copied for a fee, at the NRC's Public Document Room (PDR), located at
One White Flint North, Public File Area O1F21, 11555 Rockville Pike
(first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the 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
[[Page 17440]]
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 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
[[Page 17441]]
or party to use E-Filing if the presiding officer subsequently
determines that the reason for granting the exemption from use of E-
Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
social security numbers, home addresses, or home phone numbers in their
filings, unless an NRC regulation or other law requires submission of
such information. With respect to copyrighted works, except for limited
excerpts that serve the purpose of the adjudicatory filings and would
constitute a Fair Use application, participants are requested not to
include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have
access to ADAMS or who encounter problems in accessing the documents
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].
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 amendment request: October 29, 2009.
Description of amendment request: The amendments would delete a
license condition located in each of the unit's Facility Operating
Licenses (FOLs) which restricts the maximum fuel rod average burnup.
Deletion of this condition would allow the maximum fuel rod average
burnup to increase.
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.
Deletion of the MNS [McGuire Nuclear Station] and CNS [Catawba
Nuclear Station] FOL Appendix B conditions currently limiting
maximum rod average burnup to 60 GWd/MTU [Gigawatt-day per Metric
Ton Uranium] does not add, delete, or modify any MNS or CNS systems,
structures, or components (SSCs). The proposed amendment would
effectively allow future increases in the MNS and CNS maximum rod
average burnup limit up to and including 62 GWd/MTU using existing
fuel management methods, analyses, and models that have been
reviewed and approved by the NRC [Nuclear Regulatory Commission].
Maximum average rod burnup limits will continue to be maintained
within safe and acceptable limits using these fuel management
methods and models.
Increasing the MNS and CNS maximum rod average burnup limit does
not affect the thermal hydraulic response or the radiological
consequences of any previously evaluated accident. The fuel rod
design criteria will continue to be met at the maximum burnup limits
allowed utilizing the current fuel management, analysis, and
evaluation processes. An increase to the maximum rod average burnup
limit will not increase the likelihood of a malfunction of nuclear
fuel since the fuel currently used at MNS and CNS has been designed
to support a maximum rod average burnup up to and including 62 GWd/
MTU. 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 amendment would delete MNS and CNS FOL Appendix B
conditions which currently limits maximum rod average burnup to 60
GWd/MTU. The proposed amendment would effectively allow future
increases in the MNS and CNS maximum rod average burnup limit up to
and including 62 GWd/MTU using existing fuel management methods,
analyses, and models that have been reviewed and approved by the
NRC. The proposed amendment does not change the design function of
the nuclear fuel or create any credible new failure mechanisms or
malfunctions for the nuclear fuel. Fuel rod design criteria will
continue to be met at the maximum burnup limits allowed under the
fuel management methods and models that have been previously
reviewed and approved by the NRC. Therefore, the proposed amendment
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 would delete a MNS and CNS FOL Appendix B
conditions which currently limits maximum rod average burnup to 60
GWd/MTU. The proposed amendment would effectively allow future
increases in the MNS and CNS maximum rod average burnup limit up to
and including 62 GWd/MTU using existing fuel management methods,
analyses, and models that have been reviewed and approved by the
NRC. The proposed amendment does not result in altering or exceeding
a design basis or safety limit for the plant. All current fuel
design criteria will continue to be satisfied, and the safety
analysis of record, including evaluations of the radiological
consequences of design bases accidents, will remain applicable.
Radiological consequences have been evaluated consistent with
methodologies approved by the NRC. [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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: October 29, 2009.
Description of amendment request: The amendments would delete a
license condition located in each of the unit's Facility Operating
Licenses (FOLs) which restricts the maximum fuel rod average burnup.
Deletion of this condition would allow the maximum fuel rod average
burnup to increase.
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.
Deletion of the MNS [McGuire Nuclear Station] and CNS [Catawba
Nuclear Station]
[[Page 17442]]
FOL Appendix B conditions currently limiting maximum rod average
burnup to 60 GWd/MTU [Gigawatt-day per Metric Ton Uranium] does not
add, delete, or modify any MNS or CNS systems, structures, or
components (SSCs). The proposed amendment would effectively allow
future increases in the MNS and CNS maximum rod average burnup limit
up to and including 62 GWd/MTU using existing fuel management
methods, analyses, and models that have been reviewed and approved
by the NRC [Nuclear Regulatory Commission]. Maximum average rod
burnup limits will continue to be maintained within safe and
acceptable limits using these fuel management methods and models.
Increasing the MNS and CNS maximum rod average burnup limit does
not affect the thermal hydraulic response or the radiological
consequences of any previously evaluated accident. The fuel rod
design criteria will continue to be met at the maximum burnup limits
allowed utilizing the current fuel management, analysis, and
evaluation processes. An increase to the maximum rod average burnup
limit will not increase the likelihood of a malfunction of nuclear
fuel since the fuel currently used at MNS and CNS has been designed
to support a maximum rod average burnup up to and including 62 GWd/
MTU. 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 amendment would delete MNS and CNS FOL Appendix B
conditions which currently limits maximum rod average burnup to 60
GWd/MTU. The proposed amendment would effectively allow future
increases in the MNS and CNS maximum rod average burnup limit up to
and including 62 GWd/MTU using existing fuel management methods,
analyses, and models that have been reviewed and approved by the
NRC. The proposed amendment does not change the design function of
the nuclear fuel or create any credible new failure mechanisms or
malfunctions for the nuclear fuel. Fuel rod design criteria will
continue to be met at the maximum burnup limits allowed under the
fuel management methods and models that have been previously
reviewed and approved by the NRC. Therefore, the proposed amendment
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 would delete a MNS and CNS FOL Appendix B
conditions which currently limits maximum rod average burnup to 60
GWd/MTU. The proposed amendment would effectively allow future
increases in the MNS and CNS maximum rod average burnup limit up to
and including 62 GWd/MTU using existing fuel management methods,
analyses, and models that have been reviewed and approved by the
NRC. The proposed amendment does not result in altering or exceeding
a design basis or safety limit for the plant. All current fuel
design criteria will continue to be satisfied, and the safety
analysis of record, including evaluations of the radiological
consequences of design bases accidents, will remain applicable.
Radiological consequences have been evaluated consistent with
methodologies approved by the NRC. [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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: February 8, 2010.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) requirements related to TS 3.1.3,
``Control Rod Operability,'' and TS 3.1.5, ``Control Rod Scram
Accumulators,'' to be consistent with NUREG-1433, ``Standard Technical
Specifications General Electric Plants, BWR/4.'' The proposed amendment
also corrects certain typographical errors.
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 changes involve an administrative change to LCO
[limiting condition for operation] 3.1.3, ``Control Rod
OPERABILITY,'' and a simplification in the modeling methodology for
scram time analysis in LCO 3.1.5, ``Control Rod Scram
Accumulators,'' that continue to ensure that control rod operability
requirements for the number and distribution of operable, slow and
stuck control rods satisfy scram reactivity rate assumptions used in
the plant safety analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve any physical alteration of
the plant (no new or different type of equipment is being installed)
and do not involve a change in the design, normal configuration, or
basic operation of the plant. The proposed changes do not introduce
any new accident initiators. The proposed changes do not involve
significant changes in the fundamental methods governing normal
plant operation and do not require unusual or uncommon operator
actions. The proposed changes provide assurance that the plant will
not be operated in a mode or condition that violates the assumptions
or initial conditions in the safety analyses and that the systems,
structures, and components (SSCs) remain capable of performing their
intended safety functions as assumed in the same analyses.
Consequently, the response of the plant and the plant operator to
postulated events will not be significantly different.
Therefore, the proposed TS 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?
Response: No.
Margin of safety is related to confidence in the ability of
fission product barriers to perform their intended design functions
during and following an accident. The proposed changes address
control rod operability and continue to ensure control rod scram
time acceptance criteria is satisfied. The scram time test
acceptance criteria and control rod operability restrictions are
based on industry approved methodology and will continue to ensure
control rod scram design functions and reactivity insertion
assumptions used in the safety analyses continue to be protected.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: January 28, 2010.
Description of amendment request: The proposed license amendment
[[Page 17443]]
request modifies the licensee's commitment to Table B-1, ``Minimum
Staffing Requirements for NRC Licensees for Nuclear Power Plant
Emergencies,'' of NUREG-0654/FEMA-REP-1, Revision 1, ``Criteria for
Preparation and Evaluation of Radiological Emergency Response Plans and
Preparedness in Support of Nuclear Power Plants,'' dated November 1980.
Current Table 13.3-17, ``Repair and Corrective Actions,'' of the
Emergency Plan only allows that Electrical or Instrumentation & Control
technicians may fill these two positions. This change will allow these
two maintenance positions on shift to be filled with any combination of
the three maintenance craft disciplines.
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 not involve a significant increase in the probability
or consequences of an accident previously evaluated.
No.
The proposed change does not increase the probability or
consequences of an accident. The change only impacts the
implementation of the Emergency Plan by changing staffing of the
Repair and Corrective action functions after an event. It has no
impact on plant equipment or the operation of plant equipment and
thus has no impact on the probability or consequences of an event.
The number of personnel on shift has not been revised from the
current Emergency Plan. The repair and corrective action function
would continue to be performed by trained personnel because the
process, personnel, and equipment involved in implementing the
Emergency Plan would complete the same functions as those completed
under the existing Emergency Plan, the Plan would continue to ensure
adequate protection of public health and safety.
(2) Does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
No.
The change only impacts the implementation of the Emergency Plan
by changing staffing of the Repair and Corrective action functions
after an event. The change does not impact any plant equipment or
systems needed to respond to an accident, nor does it involve any
analysis of plant accidents. The proposed change does not create a
new or different kind of accident from any previously evaluated
because this change only impacts emergency response repair
functions.
(3) Does not involve a significant reduction in a margin of
safety.
No.
The change to the Emergency Plan does not reduce the margin of
safety currently provided by the Plan as it maintains the current
number of personnel on shift to perform Repair and Corrective action
functions. Repair and corrective actions will continue to be
performed by trained personnel. 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
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: January 24, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 1.0, Definitions, TS
Section 3.6, Primary System Boundary Specifications 3.6.A, and TS
Administrative Controls Section 5.5, to include reference to the
Pressure and Temperature Limits Report (PTLR). The PTLR includes
revised 34 effective full-power years (EFPY) P-T Curves, neutron
fluence, and Adjusted Reference Temperature (ART) values.
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 modifies Technical Specifications (TS)
Section 1.0
(``Definitions''), Specification 3.6.A.2, and revises 5.0
(``Administrative Controls''), to include section 5.5.9 to include
reference to the Pressure and Temperature Limits Report (PTLR). This
change adopts the methodology of SIR-05-044-A, ``Pressure-
Temperature Limits Report Methodology for Boiling Water Reactors,''
dated April-2007 for preparation of the pressure and temperature
curves, and incorporates the guidance of TSTF [Technical
Specification Task Force] -419-A (``Revised PTLR Definition and
References in ISTS 5.6.6, RCS [reactor coolant system] PTLR''). In
an NRC Safety Evaluation [safety evaluation] Report dated February
6, 2007, ``the NRC staff has found that SIR-05-044 is acceptable for
referencing in licensing applications for General Electric-designed
boiling water reactors to the extent,'' specified and under, the
limitations delineated in the TR and in the enclosed final SE.'' As
part of this change, the Pilgrim Pressure and Temperature Limits
Report (PTLR) based on the methodology and template provided in SIR-
05-044-A is being supplied for review. The pressure and temperature
curves utilize the methodology of SIR-05-044-A.
The NRC has established requirements in Appendix G to 10 CFR
[Part] 50 in order to protect the integrity of the reactor coolant
pressure boundary (RCPB) in nuclear power plants. Additionally, the
regulation in 10 CFR Part 50, Appendix H, provides the NRC staff's
criteria for the design and implementation of RPV material
surveillance programs for operating light water reactors.
Implementing this NRC approved methodology does not reduce the
ability to protect the reactor coolant pressure boundary as
specified in Appendix G, nor will this change increase the
probability of malfunction of plant equipment, or the failure of
plant structures, systems, or components. Incorporation of the new
methodology for calculating P-T curves, and the relocation of the P-
T curves from the TS to the PTLR provides an equivalent level of
assurance that the RCPB is capable of performing its intended safety
functions. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not affect the assumed accident
performance of the RCPB, nor any plant structure, system, or
component previously evaluated. The proposed change does not involve
the installation of new equipment, and installed equipment is not
being operated in a new or different manner. The change in
methodology ensures that the RCPB remains capable of performing its
safety functions. No set points are being changed which would alter
the dynamic response of plant equipment. Accordingly, no new failure
modes are introduced which could introduce 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 does not affect the function of the RCPB or
its response during plant transients. There are no changes proposed
which alter the set points at which protective actions are
initiated, and there is no change to the operability requirements
for equipment assumed to operate for accident mitigation. This
change adopts the methodology of SIR-05-044-A, ``Pressure-
Temperature Limits Report Methodology for Boiling Water Reactors,''
dated April 2007 for preparation of the pressure and temperature
curves. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
This change adopts the methodology of SIR-05-044-A, ``Pressure-
Temperature
[[Page 17444]]
Limits Report Methodology for Boiling Water Reactors,'' dated April
2007 for preparation of the pressure and temperature curves, and
incorporates the guidance of TSTF-419-A (``Revise PTLR Definition
and References in [Improved Standard Technical Specification] ISTS
5.6.6, RCS PTLR''). In an NRC Safety Evaluation Report dated
February 6, 2007, the NRC staff has found that SIR-05-044 is
acceptable for referencing in licensing applications for General
Electric-designed boiling water reactors.''
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.
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: December 3, 2009.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) to incorporate Standard Technical
Specification 3.1.8 ``Scram Discharge Volume (SDV) Vent and Drain
Valves'' and associated Bases of NUREG-1433, Revision 3, ``Standard
Technical Specifications General Electric Plants, BWR/4,'' modified to
account for plant specific design details.
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 operation of Vermont Yankee Nuclear Power Station (VY) in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed amendment does not impact the operability of any
structure, system or component that affects the probability of an
accident or that supports mitigation of an accident previously
evaluated. The proposed amendment does not affect reactor operations
or accident analysis and has no radiological consequences. The
operability requirements for accident mitigation systems remain
consistent with the licensing and design basis. Therefore, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The operation of VY in accordance with the proposed amendment
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
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 plant operation. Thus, this
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The operation of VY in accordance with the proposed amendment
will not involve a significant reduction in a margin of safety.
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by valves in the vent and drain lines and by the required
action to isolate the affected line. The ability to vent and drain
the SDVs is maintained through administrative controls. In addition,
the reactor protection system ensures that an SDV will not be filled
to the point that it has insufficient volume to accept a full scram.
Maintaining the safety functions related to isolation of the SDV and
insertion of control rods ensures that the proposed change does not
involve a significant reduction in the margin of safety. The
proposed amendment does not change the design or function of any
component or system. The proposed amendment does not impact any
safety limits, safety settings or safety margins. 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.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP 2), Oswego County, New York
Date of amendment request: December 9, 2009.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.8.4, ``DC Sources--Operating,''
by removing the Mode restrictions for performance of TS Surveillance
Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division 3 direct
current (DC) electrical power subsystem battery. These surveillances
verify that the battery capacity is adequate for the battery to perform
its required functions. The proposed amendment would remove these Mode
restrictions for the Division 3 battery, thereby allowing performance
of SR 3.8.4.7 and SR 3.8.4.8 for the Division 3 battery during Mode 1,
2, or 3 in conjunction with scheduled high pressure core spray (HPCS)
system outages. Eliminating the requirement to perform SR 3.8.4.7 and
SR 3.8.4.8 during Mode 4 or 5 (cold shutdown or refueling conditions)
will provide greater flexibility in scheduling Division 3 battery
testing activities by allowing the testing to be performed during non-
outage times.
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 Division 3 (HPCS) DC electrical power subsystem and its
associated emergency loads are accident mitigating features, not
accident initiators. Therefore, the proposed TS changes to allow
performance of Division 3 battery surveillance testing (service test
and the battery performance discharge test) in any plant operating
mode will not significantly impact the probability of any previously
evaluated accident.
The design and function of plant equipment is not being modified
by the proposed amendment. Neither the battery test frequency nor
the time that the TSs allow the HPCS system to be inoperable are
being revised. Battery testing in accordance with the proposed TS
changes will continue to verify that the Division 3 DC electrical
power subsystem is capable of performing its required function of
providing DC power to HPCS system equipment, consistent with the
plant safety analyses. The battery testing period is within the
period of time that the HPCS system will already be out of service
for a planned system outage. The battery testing does not increase
unavailability of the supported HPCS system or represent any change
in risk above the current practice of planned system maintenance
outages. Any risk associated with the testing of the Division 3
battery will be enveloped by the risk management of the HPCS system
outage. In addition, the HPCS system reliability and availability
are monitored and evaluated in relationship to Maintenance Rule
goals to ensure that total outage times do not degrade operational
safety over time.
Testing is limited to only one electrical division of equipment
at a time to ensure that design basis requirements are met. Should a
fault occur while testing the Division 3 battery, there would be no
significant impact on any accident consequences since the other two
divisional DC electrical power subsystems and their associated
emergency loads would be available to provide the minimum safety
functions necessary to shut
[[Page 17445]]
down the unit and maintain it in a safe shutdown condition.
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.
No changes are being made to the plant that would introduce any
new accident causal mechanisms. Equipment will be operated in the
same configuration with the exception of the plant operating mode in
which the Division 3 battery surveillance testing is conducted.
Performance of these surveillance tests while online will continue
to verify operability of the Division 3 battery. The proposed
license amendment does not impact any plant systems that are
accident initiators and does not adversely impact any accident
mitigating systems, since the HPCS system will already be out of
service. The battery testing will not increase the out-of-service
time for the HPCS system.
Therefore, the proposed amendment 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.
Margin of safety is related to confidence in the ability of the
fission product barriers (fuel cladding, reactor coolant system, and
primary containment) to perform their design functions during and
following postulated accidents. The proposed changes to the TS
surveillance testing requirements for the Division 3 battery do not
affect the operability requirements for the battery, as verification
of such operability will continue to be performed as required.
Continued verification of operability supports the capability of the
Division 3 DC electrical power subsystem to perform its required
function of providing DC power to HPCS system equipment, consistent
with the plant safety analyses. Consequently, the performance of the
fission product barriers will not be adversely impacted by
implementation of the proposed amendment. In addition, the proposed
changes do not alter setpoints or limits established or assumed by
the accident analysis.
The battery testing will be performed when the HPCS system is
already out of service for a planned system outage. The battery
testing does not increase unavailability of the supported HPCS
system or represent any change in risk above the current practice of
planned system maintenance outages, as currently allowed by the TS.
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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Nancy L. Salgado.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP 2), Oswego County, New York
Date of amendment request: December 18, 2009.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) requirements for unavailable
barriers by adding limiting condition for operation (LCO) 3.0.9. The
NRC staff issued a Notice of Opportunity to Comment in the Federal
Register on June 2, 2006 (71 FR 32145), on possible amendments to
revise the plant-specific TSs, including a model safety evaluation and
model no significant hazards consideration determination using the
consolidated line-item improvement process. The NRC staff subsequently
issued a Notice of Availability of the models for referencing in
license amendment applications in the Federal Register on October 3,
2006 (71 FR 58444). The licensee affirmed the applicability of the
model no significant hazards consideration determination in its
application dated December 18, 2009.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an unavailable barrier if risk is assessed and managed. The
postulated initiating events which may require a functional barrier are
limited to those with low frequencies of occurrence, and the overall TS
system safety function would still be available for the majority of
anticipated challenges. Therefore, the probability of an accident
previously evaluated is not significantly increased, if at all. The
consequences of an accident while relying on the allowance provided by
proposed LCO 3.0.9 are no different than the consequences of an
accident while relying on the TS required actions in effect without the
allowance provided by proposed LCO 3.0.9. Therefore, the consequences
of an accident previously evaluated are not significantly affected by
this change. The addition of a requirement to assess and manage the
risk introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to an unavailable barrier, if risk is
assessed and managed, will not introduce new failure modes or effects
and will not, in the absence of other unrelated failures, lead to an
accident whose consequences exceed the consequences of accidents
previously evaluated. The addition of a requirement to assess and
manage the risk introduced by this change will further minimize
possible concerns. Thus, this change does not create the possibility of
a new or different kind of accident from an accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an unavailable
barrier, if risk is assessed and managed. The postulated initiating
events which may require a functional barrier are limited to those with
low frequencies of occurrence, and the overall TS system safety
function would still be available for the majority of anticipated
challenges. The risk impact of the proposed TS changes was assessed
following the three-tiered approach recommended in RG [Regulatory
Guide] 1.177. A bounding risk assessment was performed to justify the
proposed TS changes. This application of LCO 3.0.9 is predicated upon
the licensee's performance of a risk assessment and the management of
plant risk. The net change to the margin of safety is insignificant as
indicated by the anticipated low levels of associated risk (ICCDP
[Incremental Conditional Core Damage Probability] and ICLERP
[Incremental Conditional Large Early Release Probability]) as shown in
Table 1 of Section 3.1.1 in the Safety Evaluation published in the
Federal Register on October 3, 2006. Therefore, this change does not
involve a
[[Page 17446]]
significant reduction in a margin of safety.
The NRC staff has reviewed the analysis and, based on this review,
it appears that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Nancy L. Salgado.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: October 27, 2009.
Description of amendment request: The proposed amendment would
adopt the Alternative Source Term (AST) methodology, in addition to
Technical Specification (TS) changes supported by the AST design basis
accident radiological consequences analyses. The proposed amendment
would also incorporate Technical Specification Task Force (TSTF)-490,
``Deletion of E-Bar Definition and Revision to RCS [reactor coolant
system] Specific Activity Tech Spec,'' Revision 0.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
With this change, Prairie Island Nuclear Generating Plant
(PINGP) proposes to implement 10 CFR 50.67, alternative source term
methodologies, implement approved industry improved Standard
Technical Specification traveler, TSTF-490, and revise TS 3.3.7,
``Spent Fuel Pool Special Ventilation System Actuation
Instrumentation,'' TS 3.7.12, ``Auxiliary Building Special
Ventilation System,''
TS 3.7.13, ``Spent Fuel Pool Special Ventilation System,'' TS
3.9.4, ``Containment Penetrations,'' TS 5.5.9, ``Ventilation Filter
Testing Program,'' TS 5.5.14, ``Containment Leakage Rate Testing
Program,'' and TS 5.5.16, ``Control Room Habitability Program.''
Alternative source term (AST) calculations have been performed
for PINGP that demonstrate the dose consequences are consistent with
the regulatory limits of 10 CFR 50.67 and the guidance of Regulatory
Guide (RG) 1.183. The use of the AST methodology changes the
regulatory assumptions regarding the analytical treatment of the
design basis accidents and has no direct effect on the probability
of any accident. AST methods have been utilized in the analysis of
the limiting design basis accidents, as follows: loss of coolant
accident, fuel handling accident, main steam line break, steam
generator tube rupture, control rod ejection accident, and locked
rotor accident. The results of the analyses, which include the
proposed changes to the Technical Specifications, demonstrate that
the dose consequences of these limiting events are within regulatory
limits.
Reactor coolant specific activity is not an initiator for any
accident previously evaluated. The Completion Time when reactor
coolant gross activity is not within limit is not an initiator for
any accident previously evaluated. The current variable limit on
primary coolant iodine concentration is not an initiator to any
accident previously evaluated. As a result, the proposed change does
not significantly increase the probability of an accident. The
proposed change will limit reactor coolant noble gases to
concentrations consistent with the accident analyses. The proposed
change to the Completion Time has no impact on the consequences of
any design basis accident since the consequences of an accident
during the extended Completion Time are the same as the consequences
of an accident during the current Completion Time. As a result, the
consequences of any accident previously evaluated are not
significantly increased.
The Spent Fuel Pool Special Ventilation System is no longer
credited for filtration or isolation. The Containment Penetrations
TS is being replaced with a TS on Decay Time, which requires that
recently irradiated fuel (<50 hours) cannot be handled. The
Ventilation Filter Testing Program TS is being revised to reflect
changes to filter testing. As a result of these TS changes, the
probability or consequences of an accident previously evaluated are
not significantly increased.
Based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
With this change, PINGP proposes to implement 10 CFR 50.67,
alternative source term methodologies, implement approved industry
improved Standard Technical Specification traveler, TSTF-490, and
revise TS 3.3.7, ``Spent Fuel Pool Special Ventilation System
Actuation Instrumentation,'' TS 3.7.12, ``Auxiliary Building Special
Ventilation System,'' TS 3.7.13, ``Spent Fuel Pool Special
Ventilation System,'' TS 3.9.4, ``Containment Penetrations,'' TS
5.5.9, ``Ventilation Filter Testing Program,'' TS 5.5.14,
``Containment Leakage Rate Testing Program,'' and TS 5.5.16,
``Control Room Habitability Program.''
The AST methodology is not an accident initiator, as it is a
method used to estimate resulting accident doses. The proposed
operation of plant systems affected by this change does not create
the possibility of a new or different kind of accident previously
evaluated. Changes that are proposed to plant equipment (ventilation
systems) pertain to accident mitigation equipment. The operation or
mis-operation of these ventilation systems do not initiate any
accidents. The radiological consequence analyses demonstrate that
the proposed changes are acceptable. The results of the analyses,
which include the proposed changes to the Technical Specifications,
demonstrate that the dose consequences of these limiting events are
within regulatory limits.
The proposed change in specific activity limits does not alter
any physical part of the plant nor does it affect any plant
operating parameter. The change does not create the potential of a
new or different kind of accident from any accident previously
evaluated.
Based on the above, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
With this change, PINGP proposes to implement 10 CFR 50.67,
alternative source term methodologies, implement approved industry
improved Standard Technical Specification traveler, TSTF-490, and
revise TS 3.3.7, ``Spent Fuel Pool Special Ventilation System
Actuation Instrumentation,'' TS 3.7.12, ``Auxiliary Building Special
Ventilation System,'' TS 3.7.13, ``Spent Fuel Pool Special
Ventilation System,'' TS 3.9.4, ``Containment Penetrations,'' TS
5.5.9, ``Ventilation Filter Testing Program,'' TS 5.5.14,
``Containment Leakage Rate Testing Program,'' and TS 5.5.16,
``Control Room Habitability Program.''
The proposed implementation of the AST methodology is consistent
with RG 1.183. The radiological consequences of these accidents are
within the regulatory acceptance criteria associated with the use of
the AST methodology. The doses at the exclusion area and low
population zone boundaries and in the control room are consistent
with the regulatory limits of 10 CFR 50.67 and the guidance of RG
1.183. The margin of safety for the radiological consequences of
these accidents is considered to be that provided by meeting the
applicable regulatory limits, which are set at or below 10 CFR 50.67
limits.
The proposed change to revise the limits on noble gas
radioactivity in the primary coolant is consistent with the
assumptions in the safety analyses and will ensure the monitored
values protect the initial assumptions in the safety analyses.
Based on the above, 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: Peter M. Glass, Assistant General Counsel,
Xcel Energy
[[Page 17447]]
Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Robert J. Pascarelli.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: January 29, 2010.
Description of amendment request: The amendments would change an
Emergency Action Level (EAL) scheme based on NUREG-0654, ``Criteria for
Preparation and Evaluation of Radiological Emergency Response Plan and
Preparedness in Support of Nuclear Power Plants,'' to one based on NEI
99-01, ``Methodology for Development of Emergency Action Levels,''
Revision 4. This would change the methodology for deriving selected
Notification of Unusual Event values in Table R-1, Gaseous Effluent
Monitor Classification Thresholds, and deleting EAL RA2.4 which
evaluates abnormal radiation readings at infrequently accessed areas
and revise the radiation level threshold values for Reactor Coolant
System (RCS) letdown indication.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1:
Does the proposed amendment involve a significant increase in
the probability or Consequences of an accident previously evaluated?
Response: No.
These changes affect the North Anna [* * *] Power Station
Emergency Action Levels, but do not alter any of the requirements of
the Operating License or the Technical Specifications. The proposed
changes do not modify any plant equipment and do not impact any
failure modes that could lead to an accident. Additionally, the
proposed changes have no effect on the consequences of any analyzed
accident since the changes do not affect any equipment related to
accident mitigation. Based on this discussion, the proposed
amendment does not increase the probability or consequence of an
accident previously evaluated.
Criterion 2:
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
These changes affect the North Anna [* * *] Power Station
Emergency Action Levels, but do not alter any of the requirements of
the Operating License or the Technical Specifications. They do not
modify any plant equipment and there is no impact on the capability
of the existing equipment to perform their intended functions. No
system setpoints are being modified. No new failure modes are
introduced by the proposed changes. The proposed amendment does not
introduce accident initiator or malfunctions that would cause a new
or different kind of accident. Therefore, the proposed amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3:
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
These changes affect the North Anna [* * *] Power Station
Emergency Action Levels, but do not alter any of the requirements of
the Operating License or the Technical Specifications. The proposed
changes do not affect any of the assumptions used in the accident
analysis, nor do they affect any operability requirements for
equipment important to plant safety. Therefore, the proposed changes
will not result in a significant reduction in the margin of safety
as defined in the bases for technical specifications covered in this
license amendment request. [Therefore, this change does not involve
a significant reduction in a margin of safety.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: Gloria Kulesa.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: January 27, 2010.
Description of amendment request: The proposed license amendment
request would increase each unit's rated power (RP) level from 2546
megawatts thermal (MWt) to 2587 MWt, and make Technical Specifications
changes as necessary to support operation at the uprated power level.
The proposed change is an increase in RP of approximately 1.6%.
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 consequence of an accident previously evaluated?
Response: No.
The proposed change will increase the Surry Power Station (SPS)
Units 1 and 2 rated power (RP) from 2546 megawatts thermal (MWt) to
2587 MWt. Nuclear steam supply system and balance-of-plant systems,
components and analyses that could be affected by the proposed
change to the RP were evaluated using revised design parameters. The
evaluations determined that these structures, systems and components
are capable of performing their design function at the proposed
uprated RP of 2587 MWt. An evaluation of the accident analyses
demonstrates that the applicable analysis acceptance criteria are
still met with the proposed changes. Power level is an input
assumption to equipment design and accident analyses, but it is not
a transient or accident initiator. Accident initiators are not
affected by the power uprate, and plant safety barrier challenges
are not created by the proposed changes.
The radiological consequences of operation at the uprated power
conditions have been assessed. The proposed change to RP does not
affect release paths, frequency of release, or the analyzed reactor
core fission product inventory for any accidents previously
evaluated in the SPS Updated Final Safety Analysis Report. There is
a small increase in the reactor coolant activity concentration.
Structures, systems and components required to mitigate transients
are capable of performing their design functions with the proposed
changes, and are thus acceptable. Analyses performed to assess the
effects of mass and energy releases remain valid. The assessment of
radiological consequences for operation at the proposed power level
confirmed that there is not a significant increase for affected
events.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of any proposed changes. The
ultrasonic flow meter (UFM) being installed to facilitate the
Measurement Uncertainty Recapture (MUR) power uprate has been
analyzed, and system failures will not adversely affect any safety-
related system or any structures, systems or components required for
transient mitigation. Structures, systems and components previously
required for transient mitigation are still capable of fulfilling
their intended design functions. The proposed changes have no
significant adverse affect on any safety-related structures, systems
or components and do not significantly change the performance or
integrity of any safety-related system.
The proposed changes do not adversely affect any current system
interfaces or create any new interfaces that could result in an
accident or malfunction of a different kind than previously
evaluated. Operating at an RP of 2587 MWt does not create any new
accident initiators or precursors. Credible
[[Page 17448]]
malfunctions are bounded by the current accident analyses of record
or recent evaluations demonstrating that applicable criteria are
still met with the proposed changes.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margins of safety associated with the power uprate are those
pertaining to core thermal power. These include fuel cladding,
reactor coolant system pressure boundary, and containment barriers.
Core analyses demonstrate that power uprate implementation will
continue to meet the current nuclear design basis. Impacts to
components associated with the reactor coolant system pressure
boundary structural integrity, and factors such as pressure-
temperature limits, vessel fluence, and pressurized thermal shock
were determined to be bounded by the current analyses.
Systems will continue to operate within their design parameters
and remain capable of performing their intended safety functions
following implementation of the proposed change. The current SPS
safety analyses, and the revised design basis radiological accident
dose calculations, bound the power uprate without significantly
impacting margins.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA
23219.
NRC Branch Chief: Gloria Kulesa.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: January 29, 2010.
Description of amendment request: The amendments would change an
Emergency Action Level (EAL) scheme based on NUREG-0654, ``Criteria for
Preparation and Evaluation of Radiological Emergency Response Plan and
Preparedness in Support of Nuclear Power Plants,'' to one based on NEI
99-01, ``Methodology for Development of Emergency Action Levels,''
Revision 4. This would change the methodology for deriving selected
Notification of Unusual Event values in Table R-1, Gaseous Effluent
Monitor Classification Thresholds, and deleting EAL RA2.4 which
evaluates abnormal radiation readings at infrequently accessed areas.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
These changes affect the [* * *] Surry Power Station Emergency
Action Levels, but do not alter any of the requirements of the
Operating License or the Technical Specifications. The proposed
changes do not modify any plant equipment and do not impact any
failure modes that could lead to an accident. Additionally, the
proposed changes have no effect on the consequences of any analyzed
accident since the changes do not affect any equipment related to
accident mitigation. Based on this discussion, the proposed
amendment does not increase the probability or consequence of an
accident previously evaluated.
Criterion 2:
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
These changes affect the [* * *] Surry Power Station Emergency
Action Levels, but do not alter any of the requirements of the
Operating License or the Technical Specifications. They do not
modify any plant equipment and there is no impact on the capability
of the existing equipment to perform their intended functions. No
system setpoints are being modified. No new failure modes are
introduced by the proposed changes. The proposed amendment does not
introduce accident initiator or malfunctions that would cause a new
or different kind of accident. Therefore, the proposed amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3:
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
These changes affect [* * *] the Surry Power Station Emergency
Action Levels, but do not alter any of the requirements of the
Operating License or the Technical Specifications. The proposed
changes do not affect any of the assumptions used in the accident
analysis, nor do they affect any operability requirements for
equipment important to plant safety. Therefore, the proposed changes
will not result in a significant reduction in the margin of safety
as defined in the bases for technical specifications covered in this
license amendment request. [Therefore, this change does not involve
a significant reduction in a margin of safety.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Gloria Kulesa.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 16, 2009.
Description of amendment request: The proposed changes would revise
Technical Specification (TS) 3.8.4, ``DC [Direct Current] Sources--
Operating,'' Surveillance Requirement (SR) 3.8.4.2 and SR 3.8.4.5 to
revise the battery connection resistance acceptance criteria for inter-
cell connections from <= 150E-6 ohms to <= 33E-6 ohms and would add
connection resistance acceptance criteria for inter-tier connections
and inter-bank connection of <= 150E-6 ohms.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No
The proposed changes to revise the SR 3.8.4.2 and SR 3.8.4.5
acceptance criteria for battery connection resistance will not
challenge the ability of the safety-related batteries to perform
their safety function. Appropriate monitoring and maintenance will
continue to be performed on the safety related batteries. Current TS
testing and monitoring requirements will not be altered.
The proposed change does not involve a physical change to the
batteries, nor does it change the safety function of the batteries.
The proposed TS revision involves no significant changes to the
operation of any systems or components in normal and accident
operating conditions and no changes to existing structures, systems
or components.
Therefore, this change will not increase the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No
The proposed changes to revise the SR 3.8.4.2 and SR 3.8.4.5
acceptance criteria for battery connection resistance is an increase
in conservatism, without a change in system
[[Page 17449]]
testing methods, operation, or control. Safety related batteries
installed in the plant will be required to meet criteria more
restrictive and conservative than current acceptance criteria and
standards. The proposed change does not affect the manner in which
the batteries are tested and maintained, thus there are no new
failure mechanisms for the system.
Therefore, this change will 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 margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes will not adversely affect
operation of plant equipment, as the changes being made are more
restrictive. These changes will not result in a change to the
setpoints at which protective actions are initiated. Sufficient DC
capacity to support operation of mitigation equipment is ensured.
The changes associated with the new battery maintenance and
monitoring program will ensure that the station batteries are
maintained in a highly reliable manner. The equipment fed by the DC
electrical sources will continue to provide adequate power to safety
related loads in accordance with analysis assumptions.
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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, N.W., 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.
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 amendment request: October 2, 2008.
Brief description of amendment request: The proposed amendment
would revise the Technical Specifications (TS) associated with the
verification of ice condenser door operability and TS surveillance
requirements 3.6.13.5 and 3.6.13.6.
Date of publication of individual notice in Federal Register: March
8, 2010 (75 FR 10513).
Expiration date of individual notice: Comments April 7, 2010;
Hearing May 7, 2010.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: October 2, 2008.
Brief description of amendment request: The proposed amendment
would revise the Technical Specifications (TS) associated with the
verification of ice condenser door operability and TS surveillance
requirements 3.6.13.5 and 3.6.13.6.
Date of publication of individual notice in Federal Register: March
8, 2010 (75 FR 10508).
Expiration date of individual notice: Comments April 7, 2010;
Hearing May 7, 2010.
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 email to [email protected].
Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: June 19, 2009, as supplemented
by letter dated October 20, 2009.
Brief description of amendment: The proposed amendment would revise
Technical Specification 3.3.1, ``Reactor Protection System
Instrumentation.'' The proposed change revises the requirements related
to the reactor protection system interlock for the turbine trip input
to the reactor protection system.
Date of issuance: March 17, 2010.
Effective date: Effective as of the date of issuance and shall be
implemented by the end of Refueling Outage 26.
Amendment No.: 222.
Renewed Facility Operating License No. DPR-23: The amendment
revises the technical specifications.
Date of initial notice in Federal Register: January 5, 2010 (75 FR
460).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated March 17, 2010.
[[Page 17450]]
Public comments received as to proposed no significant hazards
consideration (NSHC): No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1 (ANO-1), Pope County, Arkansas
Date of amendment request: March 13, 2008, as supplemented by
letter dated February 28, 2010.
Brief description of amendment: The amendment replaced the current
ANO-1 Technical Specification 3.4.12, ``RCS [Reactor Coolant System]
Specific Activity,'' limit on RCS gross specific activity with a new
limit on RCS noble gas specific activity. The noble gas specific
activity limit would be based on a new dose equivalent Xe-133
definition that would replace the current E Bar average disintegration
energy definition. In addition, the current dose equivalent I-131
definition would be revised to allow the use of additional thyroid dose
conversion factors.
Date of issuance: March 18, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 243.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications/license.
Date of initial notice in Federal Register: May 6, 2008 (73 FR
25038). The supplemental letter dated February 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 March 18, 2010.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: March 2, 2009, as supplemented
by letter dated June 24, 2009.
Brief description of amendment: The amendment modified Technical
Specification (TS) 3.3.1.1, ``Reactor Protective Instrumentation,'' and
TS 3.3.2.1, ``Engineered Safety Feature Actuation System
Instrumentation,'' specifically, Table 3.3-1, Table 4.3-1, and Table
3.3-3, to adopt a mode of applicability for the Logarithmic Power
Level--High, Pressurizer Pressure--Low, Steam Generator [SG] Pressure--
Low, and the SG Differential Pressure and Level Low functions. These
changes are consistent with NUREG-1432, Revision 3.0, ``Standard
Technical Specifications, Combustion Engineering Plants,'' dated June
2004.
Date of issuance: March 11, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 289.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: June 2, 2009 (74 FR
26433). The supplemental letter dated June 24, 2009, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register on June 2, 2009 (74
FR 26433).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 11, 2010.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 19, 2009.
Brief description of amendment: The amendment relocated the
Waterford 3 Steam Generator Level--High trip requirements from
Technical Specification Sections 2.2 and 3/4.3.1 to the Technical
Requirements Manual (TRM). This change is consistent with Technical
Specification Task Force (TSTF) 410-A, ``Relocation of Steam Generator
Level--High Trip to the TRM,'' and Revision 3 of NUREG-1432, ``Standard
Technical Specifications, Combustion Engineering Plants.''
Date of issuance: March 18, 2010.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 225.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 1, 2009 (74 FR
62834).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 18, 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 4, 2008, as
supplemented by letters dated February 17, 2009; July 27, 2009;
December 4, 2009; and January 29, 2010.
Brief description of amendment: The amendments revise Technical
Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16, ``RCS [Reactor
Coolant System] Specific Activity,'' and Surveillance Requirements
3.4.16.1, 3.4.16.2, and 3.4.16.3. The revisions replace the current TS
3.4.16 limit on RCS gross specific activity with a new limit on RCS
noble gas-specific activity. The revisions adopt TS Task Force (TSTF)
Change Traveler, TSTF-490, ``Deletion of E Bar Definition and Revision
to RCS Specific Activity Tech Spec [sic],'' Revision 0.
Date of issuance: March 23, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: Braidwood Unit 1-162; Braidwood Unit 2-162; Byron
Unit No. 1-167; and Byron Unit No. 2-167.
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: January 27, 2009 (74 FR
4771).
The supplemental letters provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 23, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: March 26, 2009, as supplemented
by letter dated October 28, 2009.
Brief description of amendments: The proposed changes would revise
Technical Specification 3.5.1, ``Emergency Core Cooling Systems (ECCS)
Operating,'' to delete the existing allowance with the Automatic
Depressurization System accumulator backup compressed gas system that
[[Page 17451]]
currently allows a completion time of 72 hours to restore bottle
pressure to >= 500 psig.
Date of issuance: March 19, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 196/183.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: September 8, 2009 (74
FR 46242). The October 28, 2009 supplement, contained clarifying
information and did not change the NRC staff's initial proposed finding
of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: November 6, 2008, supplemented
by letters dated December 11, 2008, July 2, 2009, October 2, 2009, and
November 24, 2009.
Brief description of amendment: The amendment replaces the current
TMI-1 technical specification limit on Reactor Coolant System (RCS)
gross specific activity with a new limit on RCS noble gas specific
activity. The noble gas specific activity limit is based on a new dose
equivalent Xenon-133 definition that replaces the previous E-Bar
average disintegration energy definition. In addition, the dose
equivalent Iodine-131 definition has been revised.
Date of issuance: March 11, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 272.
Renewed Facility Operating License No. DPR-50. Amendment revised
the license and the technical specifications.
Date of initial notice in Federal Register: March 10, 2009 (74 FR
10309). The supplements dated December 11, 2008, July 2, 2009, October
2, 2009, and November 24, 2009, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC staff's original
proposed no significant hazards determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 11, 2010.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 11, 2009, as supplemented by
letters dated August 12 and December 21, 2009, and March 5, 2010.
Brief description of amendment: The amendment revised Surveillance
Requirements 3.8.4.2 and 3.8.4.5 in Technical Specification Section
3.8.4, ``DC [Direct Current] Sources--Operating,'' by adding a
parameter of total battery resistance to the values of battery
connection resistance.
Date of issuance: March 18, 2010.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 236.
Facility Operating License No. DPR-46: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 5, 2009 (74 FR
20752). The supplemental letters dated August 12 and December 21, 2009,
and March 5, 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 March 18, 2010.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2 (NMP2), Oswego County, New York
Date of application for amendment: March 30, 2009, as supplemented
on November 2, 2009.
Brief description of amendment: The amendment modifies the NMP2
Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to
remove operating mode restrictions for the performance of certain
Surveillance Requirements (SRs) pertaining to the Division 3, High
Pressure Core Spray (HPCS) Emergency Diesel Generator (DG). The testing
in Modes 1 or 2 were previously prohibited in SR 3.8.1.7, SR 3.8.1.8,
and SR 3.8.1.10, and in Modes 1, 2, or 3 in SR 3.8.1.9, SR 3.8.1.11, SR
3.8.1.14, SR 3.8.1.15, and SR 3.8.1.17. The amendment removes these
Mode restrictions and allows the above SRs to be performed in any
operating mode for the Division 3 DG. The Mode restrictions remain
applicable to the other two safety-related (Division 1 and Division 2)
DGs.
Date of issuance: March 18, 2010.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 133.
Renewed Facility Operating License No. NPF-069: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: June 16, 2009 (74 FR
28577).
The supplemental letter dated November 2, 2009, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 18, 2010.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: July 27, 2009.
Description of amendment request: The amendments revised the
Technical Specifications to change Surveillance Requirement 3.6.1.3,
``Primary Containment Isolation Valves,'' to eliminate unnecessary
local leak rate tests.
Date of issuance: March 22, 2010.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 277, 304, and 263.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 20, 2009 (74 FR
53781).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 22, 2010.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: March 20, 2009, as supplemented
by letters dated December 10, 2009, and January 19, 2010.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.16, ``Containment Leakage Rate Testing
Program.'' The revision reflects a one-time extension of
[[Page 17452]]
the current containment Type A leak rate test (integrated leak rate
test or ILRT) interval requirement of Title 10 of the Code of Federal
Regulations (10 CFR) Part 50, Appendix J, ``Primary Reactor Containment
Leakage Testing for Water-Cooled Power Reactors,'' Option B,
``Performance Based Requirements,'' from 10 years to 15 years. The
amendment allows the next ILRT to be performed no later than October
25, 2014.
Date of issuance: March 17, 2010.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 195.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 25, 2009 (74 FR
42931). The supplemental letters dated December 10, 2009, and January
19, 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 March 17, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 25th day of March 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-7451 Filed 4-5-10; 8:45 am]
BILLING CODE 7590-01-P