[Federal Register Volume 75, Number 143 (Tuesday, July 27, 2010)]
[Notices]
[Pages 44020-44028]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-18078]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0256]
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 July 1, 2010 to July 14, 2010. The last
biweekly notice was published on July 13, 2010 (75 FR 39975).
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, Rules,
Announcements, and Directives Branch (RADB), TWB-05-B01M, Division of
Administrative Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be faxed to the RADB at 301-492-3446.
Documents may be examined, and/or copied for a fee, at the NRC's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
[[Page 44021]]
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
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
[[Page 44022]]
system. To be timely, an electronic filing must be submitted to the E-
Filing system no later than 11:59 p.m. Eastern Time on the due date.
Upon receipt of a transmission, the E-Filing system time-stamps the
document and sends the submitter an e-mail notice confirming receipt of
the document. The E-Filing system also distributes an e-mail notice
that provides access to the document to the NRC Office of the General
Counsel and any others who have advised the Office of the Secretary
that they wish to participate in the proceeding, so that the filer need
not serve the documents on those participants separately. Therefore,
applicants and other participants (or their counsel or representative)
must apply for and receive a digital ID certificate before a hearing
request/petition to intervene is filed so that they can obtain access
to the document via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants.
Filing is considered complete by first-class mail as of the time of
deposit in the mail, or by courier, express mail, or expedited delivery
service upon depositing the document with the provider of the service.
A presiding officer, having granted an exemption request from using E-
Filing, may require a participant or party to use E-Filing if the
presiding officer subsequently determines that the reason for granting
the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
social security numbers, home 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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: April 8, 2010.
Description of amendment request: The amendments would revise
Technical Specification (TS) 2.2, ``Safety Limit Violations,''
consistent with Technical Specification Task Force (TSTF) change
traveler TSTF-5-A, and TS 5.2.1, ``Onsite and Offsite Organizations,''
consistent with TSTF-65-A, Revision 1. Specifically, the proposed
amendment would delete redundant reporting and operational restriction
provisions from TS 2.2 and replace plant-specific organization titles
with generic organization titles in TS 5.2.1. Both TSTF-5-A and TSTF-
65-A were incorporated in Revision 2 of NUREG-1432, ``Standard
Technical Specifications for Combustion Engineering Plants.''
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.
These changes involve minor changes in organization titles and
remove redundant and unnecessary reporting requirements. The changes
are consistent with TSTF-5 and TSTF-65, which have been approved by the
NRC Staff, and included in Revision 2 of NUREG-1432. Technical
Specification Safety Limit violation reporting is redundant to 10 CFR
50.36(c)(7) and (8) and 10 CFR 50.72 and 73. The removal of the
notification, reporting, and startup requirements from the TS is an
administrative change because the current requirements duplicate what
is already contained in the regulations. The proposed changes do not
alter existing controls on plant operation (i.e., safety limit values,
LCOs [Limiting Conditions for Operations], Surveillance Requirements or
Design Features), but only remove the administrative burden of
maintaining redundant notification, reporting, and plant startup
requirements.
Functions which are necessary to operate the facility safely and in
accordance with the operating licenses remain within the organization
and will not affect the safe operation of the plant and will continue
to ensure proper control of administrative activities. The proposed
changes will not affect the operation of structures, systems, or
components, and will not reduce programmatic controls such that plant
safety would be affected.
Based on the above, the proposed amendment does not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
[[Page 44023]]
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will not affect the operation of structures,
systems, or components, and will not reduce programmatic controls such
that plant safety would be affected. The generic title changes and
deletion of redundant reporting are administrative.
Based on the above, 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 changes are administrative and will not diminish any
organizational or administrative controls currently in place. The
proposed change will not affect the operation of structures, systems,
or components, and will not reduce programmatic controls such that
plant safety would be affected. 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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: April 29, 2010.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) to incorporate Technical Specifications
Task Force (TSTF) change traveler TSTF-479-A, ``Changes to Reflect
Revision of 10 CFR 50.55a,'' as modified by TSTF-497-A, ``Limit
Inservice Testing Program SR [Surveillance Requirement] 3.0.2
Application to Frequencies of 2 Years or Less.'' Specifically, the
changes associated with TSTF-479-A would modify the reference in TS
5.5.8, ``Inservice Testing Program,'' from the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to the ASME
Code for Operation and Maintenance of Nuclear Power Plants (OM Code)
and would specify that the extension allowance of SRs is applicable to
the frequencies in the Inservice Testing Program (IST). The changes
associated with TSTF-497-A would limit the applicability of SR 3.0.2 to
frequencies of 2 years or less. In addition, the amendment would remove
the reference to component supports for consistency with the Standard
Technical Specifications because the supports are included in the
licensee's Inservice Inspection Program.
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 revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4)
regarding the IST of pumps and valves and eliminates a statement
regarding the testing of supports. The proposed changes incorporate
revisions to the ASME Code that result in a net improvement in the
measures for testing pumps and valves and the editorial change
eliminates confusion as to the testing program for supports and will
align the PVNGS specification wording to that of NUREG-1432, Revision
3.1, Standard Technical Specifications Combustion Engineering Plants.
The proposed changes do not impact any accident initiators or analyzed
events or assumed mitigation of accident or transient events, nor does
it involve the addition or removal of any equipment, or any design
changes to the facility.
Therefore, the proposed change does not represent a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4)
regarding the IST of pumps and valves and eliminates a statement
regarding the testing of supports. The proposed change incorporates
revisions to the ASME Code that result in a net improvement in the
measures for testing pumps and valves and the editorial change
eliminates confusion as to the testing program for supports and aligns
wording to that of the standard specification.
The proposed changes do not involve a modification to the physical
configuration of the plant (i.e., no new equipment will be installed)
or change in the methods governing normal plant operation. The proposed
changes will not impose any new or different requirements or introduce
a new accident initiator, accident precursor, or malfunction mechanism.
Additionally, there is no change in the types or increases in the
amounts of any effluent that may be released off-site and there is no
increase in individual or cumulative occupational exposure.
Therefore, this proposed change does not create the possibility of
an accident of a different kind than previously evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4)
regarding the inservice testing of pumps and valves and eliminates a
statement regarding the testing of supports. The proposed changes
incorporate revisions to the ASME Code that result in a net improvement
in the measures for testing pumps and valves and the editorial change
eliminates confusion as to the testing program for supports and aligns
wording to that of the standard specification. The safety functions of
the affected pumps and valves will be maintained.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034. NRC Branch Chief: Michael T. Markley.
[[Page 44024]]
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: April 29, 2010.
Description of amendment request: The amendments would remove the
Main Steam and Main Feedwater Valve Isolation Times from the Technical
Specifications (TSs) in accordance with Nuclear Regulatory Commission
(NRC)-approved TS Task Force (TSTF) Standard Technical Specification
change traveler TSTF-491, Revision 2, ``Removal of the Main Steam and
Main Feedwater Valve Isolation Times from Technical Specifications.''
The isolation times would be located outside of the TSs in a document
subject to control by the 10 CFR 50.59 process.
The NRC staff issued a Notice of Availability of ``Technical
Specification Improvement to Remove the Main Steam and Main Feedwater
Valve Isolation Time from Technical Specifications Using the
Consolidated Line Item Improvement Process,'' associated with TSTF-491,
Revision 2, in the Federal Register on December 29, 2006 (71 FR 78472).
The notice included a model license amendment request. The notice also
announced that the previously published (71 FR 193, October 5, 2006)
model safety evaluation and model No Significant Hazards Consideration
(NSHC) determination may be referenced in plant-specific applications
to adopt the changes. In its application dated April 29, 2010, the
licensee affirmed the applicability of the model NSHC determination
which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee 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 relocating main steam and main
feedwater valve isolation times to the Licensee Controlled Document
that is referenced in the Bases. The proposed change is described in
Technical Specification Task Force (TSTF) Standard TS Change
Traveler TSTF-491 related to relocating the main steam and main
feedwater valves isolation times to the Licensee Controlled Document
that is referenced in the Bases and replacing the isolation time
with the phase, ``within limits.''
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
The proposed changes relocate the main steam and main feedwater
isolation valve times to the Licensee Controlled Document that is
referenced in the Bases. The requirements to perform the testing of
these isolation valves are retained in the TS. Future changes to the
Bases or licensee-controlled document will be evaluated pursuant to
the requirements of 10 CFR 50.59, ''``Changes, test and
experiments'', to ensure that such changes do not result in more
than minimal increase in the probability or consequences of an
accident previously evaluated.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological consequences of any accident previously
evaluated. Further, the proposed changes do not increase the types
and the amounts of radioactive effluent that may be released, nor
significantly increase individual or cumulative occupation/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any 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 changes relocate the main steam and main feedwater
valve isolation times to the Licensee Controlled Document that is
referenced in the Bases. In addition, the valve isolation times are
replaced in the TS with the phase ``within limits''. The changes do
not involve a physical altering of the plant (i.e., no new or
different type of equipment will be installed) or a change in
methods governing normal [plant] operation. The requirements in the
TS continue to require testing of the main steam and main feedwater
isolation valves to ensure the proper functioning of these isolation
valves.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed changes relocate the main steam and main feedwater
valve isolation times to the Licensee Controlled Document that is
referenced in the Bases. In addition, the valve isolation times are
replaced in the TS with the phase ``within limits''. Instituting the
proposed changes will continue to ensure the testing of main steam
and main feedwater isolation valves. Changes to the Bases or license
controlled document are performed in accordance with 10 CFR 50.59.
This approach provides an effective level of regulatory control and
ensures that main steam and feedwater isolation valve testing is
conducted such that there is no significant reduction in the margin
of safety.
The margin of safety provided by the isolation valves is
unaffected by the proposed changes since there continue to be TS
requirements to ensure the testing of main steam and main feedwater
isolation valves. The proposed changes maintain sufficient controls
to preserve the current margins of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on that review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the request for amendments involves NSHC.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: June 17, 2010.
Description of amendment request: The proposed change would revise
Technical Specification (TS) 6.5.16, ``Containment Leakage Rate Testing
Program,'' to allow for the extension of the 10-year frequency of the
Arkansas Nuclear One, Unit 2 (ANO-2) Type A or Integrated Leak Rate
Test (ILRT) to be extended to 15 years on a permanent basis.
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 amendment involves changes to the ANO-2 Containment
Leakage Rate Testing Program. The proposed amendment does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. The primary containment
function is to provide an essentially leak tight barrier against the
uncontrolled release of radioactivity to the environment for
postulated accidents. As such, the containment itself and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident, do not involve any accident precursors
or initiators.
Therefore, the probability of occurrence of an accident
previously evaluated is not significantly increased by the proposed
amendment.
[[Page 44025]]
The proposed amendment adopts the NRC-accepted guidelines of
[Nuclear Energy Institute (NEI)] 94-01, Revision 2-A [``Industry
Guideline for Implementing Performance-Based Option of 10 CFR Part
50, Appendix J,'' dated October 2008], for development of the ANO-2
performance-based testing program. Implementation of these
guidelines continues to provide adequate assurance that during
design basis accidents, the primary containment and its components
will limit leakage rates to less the values assumed in the plant
safety analyses. The potential consequences of extending the ILRT
interval to 15 years have been evaluated by analyzing the resulting
changes in risk. The increase in risk in terms of person-rem
[roentgen equivalent man] per year within 50 miles resulting from
design basis accidents was estimated to be acceptably small and
determined to be within the guidelines published in [NRC Regulatory
Guide] 1.174 [``An Approach for Using Probabilistic Risk Assessment
in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis'']. Additionally, the proposed change maintains
defense-in-depth by preserving a reasonable balance among prevention
of core damage, prevention of containment failure, and consequence
mitigation. ANO-2 has determined that the increase in Conditional
Containment Failure Probability due to the proposed change would be
very small.
Therefore, it is concluded that the proposed amendment does not
significantly increase the consequences of an accident previously
evaluated.
Based on the above discussion, it is concluded that 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 amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 2-A, for the development of the ANO-2 performance-
based leakage testing program, and establishes a 15-year interval
for the performance of the containment ILRT. The containment and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident, do not involve any accident precursors
or initiators. The proposed change does not involve a physical
change to the plant (i.e., no new or different type of equipment
will be installed) or a change to the manner in which the plant is
operated or controlled.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 2-A, for the development of the ANO-2 performance-
based leakage testing program, and establishes a 15 year interval
for the performance of the containment ILRT. This amendment does not
alter the manner in which safety limits, limiting safety system
setpoints, or limiting conditions for operation are determined. The
specific requirements and conditions of the Containment Leakage Rate
Testing Program, as defined in the TS, ensure that the degree of
primary containment structural integrity and leak-tightness that is
considered in the plant's safety analysis is maintained. The overall
containment leakage rate limit specified by the TS is maintained,
and the Type A, Type B, and Type C containment leakage tests will be
performed at the frequencies established in accordance with the NRC-
accepted guidelines of NEI 94-01, Revision 2-A.
Containment inspections performed in accordance with other plant
programs serve to provide a high degree of assurance that the
containment will not degrade in a manner that is not detectable by
an ILRT. A risk assessment using the current ANO-2 PSA
[Probabilistic Safety Assessment] model concluded that extending the
ILRT test interval from ten years to 15 years results in a very
small change to the ANO-2 risk profile.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: June 14, 2010.
Description of amendment request: The proposed amendments would
allow a revision of the licensing basis, as described in the Final
Safety Analysis Report Update (FSARU), to include damping values for
the seismic design and analysis of the integrated head assembly (IHA)
that are consistent with the recommendations of Regulatory Guide (RG)
1.61, ``Damping Values for Seismic Design of Nuclear Power Plants,''
Revision 1. In addition, the RG 1.61, Revision 1, Table 1 note allowing
the use of a ``weighted average'' for design-basis safe-shutdown
earthquake (SSE) damping values applicable to steel structures of
different connection types will also be applied to determine the IHA
design-basis operating-basis earthquake (OBE) damping 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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change would allow use of critical damping values
consistent with the recommendations of RG 1.61, ``Damping Values for
Seismic Design of Nuclear Power Plants,'' Revision 1, dated March
2007, for the seismic design and analysis of the IHA. The RG 1.61,
Revision 1, Table 1 note allowing use of a ``weighted average'' for
design-basis SSE damping values applicable to steel structures of
different connection types, is also applied to determine the IHA
design-basis OBE damping values. RG 1.61, Revision 1, Table 2 for
OBE damping values does not contain the same note as found in Table
1. However use of the note for the determination of the DE [design
earthquake] damping value is consistent with the use of the note for
the determination of the DDE [double design earthquake] and HE
[Hosgri earthquake] damping values, and a weighted average more
realistically represents the IHA structure.
RG 1.61, Revision 1, specifies the damping values that the NRC
staff currently considers acceptable for complying with the agency's
regulations and guidance for seismic analysis. Revision 1
incorporates the latest data and information, and reduces
unnecessary conservatism in specification of damping values for
seismic design and analysis of SSCs [structures, systems, and
components].
The proposed change does not change the design functions of the
IHA or its response to design-basis events, nor does it affect the
capability of related SSCs to perform their design or safety
functions. The use of the proposed damping values in the seismic
design and analysis of the IHA is related to the ability of the IHA
to function in response to design-basis seismic events, and is
unrelated to the probability of occurrence of those events, or other
previously evaluated accidents. Therefore the proposed change will
not have any impact on the probability of an accident previously
evaluated.
The proposed damping values are an element of the seismic
analyses performed to confirm the ability of the IHA to function
under postulated seismic events while maintaining resulting stresses
within ASME [American Society of Mechanical Engineers Boiler and
Pressure Vessel Code] Section III allowable values. Therefore, the
use of damping values consistent with the recommendations of RG
1.61, Revision 1 does not result in an increase in the consequences
of accidents previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the
[[Page 44026]]
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve changes to any plant SSCs,
nor does it involve changes to any plant operating practice or
procedure. The damping values are an element of the seismic analyses
performed to confirm the ability of the IHA to function under
postulated seismic events while maintaining resulting stresses
within ASME Section III allowable values. Therefore, no credible new
failure mechanisms, malfunctions, or accident initiators not
considered in the design and licensing bases are created that would
create the possibility of a new or different kind of accident.
Therefore the proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The design basis of the plant requires structures to be capable
of withstanding normal and accident loads including those from a
design basis earthquake. The proposed change would allow the use of
damping values in the IHA seismic analyses that are in general more
realistic and, thus, more accurate than the damping values
recommended in RG 1.61, Revision 0, used in the analysis for the HE,
or the plant specific damping values used in the original analysis
for the DE and DDE. The NRC stated, in NUREG-0675, ``Safety
Evaluation Report Related to the Operation of Diablo Canyon Nuclear
Power Plant, Units 1 and 2,'' Supplement No. 7, that allowing use of
the higher damping values in RG 1.61, Revision 0 for the HE re-
evaluation, versus the lower values used in the original analysis,
is realistic and should not be regarded as an arbitrary lowering of
the margins of safety. The damping values in RG 1.61, Revision 0,
were based on limited data, expert opinion, and other information
available in 1973. NRC and industry research since 1973 show that
the damping values provided in the original version of RG 1.61 may
not reflect realistic damping values for SSCs. RG 1.61, Revision 1,
therefore, provides damping values based on the updated research
results that predict and estimate damping values for seismic design
of SSCs in nuclear power plants, and similarly should not be
regarded as an arbitrary lowering of the margins of safety.
As discussed above, damping values are an element of the seismic
analyses performed to confirm the ability of the IHA to function
during design-basis seismic events while maintaining resulting
stresses within ASME Section III allowable values. The proposed
change [to] allow use of damping values consistent with the
recommendations of RG 1.61, Revision 1, versus the damping values in
the current licensing basis could result in lower calculated
stresses. The analysis done for the IHA using the proposed damping
values showed the ASME Section III allowable values are met.
Sufficient safety margins are maintained when Codes and standards or
alternatives approved for use by the NRC are met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Michael T. Markley.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: May 28, 2010.
Description of amendment request: To revise Technical Specification
(TS) 4.2.2 ``Control Rod Assemblies.'' The proposed change would
include silver-indium-cadmium material in addition to the boron carbide
control rod material.
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.
Watts Bar Unit 1 Technical Specification 4.2.2, Control Rod
Assemblies, is revised to include [silver-indium-cadmium] Ag-In-Cd
material in addition to the [boron carbide] B4C control rod
material. In addition to the absorber material change, the
replacement [enhanced performance] EP Ag-In-Cd [rod cluster control
assemblies] RCCAs will be coupled with Control Rod Drive Mechanism
(CRDM) drive rod shafts which are lighter than the CRDM drive rod
shaft coupled to the B4C drive rod shafts. Also, the EP Ag-In-Cd
RCCAs are heavier than the B4C RCCAs and have a different
reactivity, or rod worth.
There are a number of events that are related to inadvertent
movement of the RCCAs; however, they are not initiated by the RCCAs.
They are initiated by the failure of plant structures, systems, or
components (SSC) other than the RCCAs. The proposed changes to the
RCCA design do not have a detrimental impact on the integrity of any
plant SSC that initiates an analyzed event. In addition, the EP Ag-
In-Cd RCCAs have the capability to mitigate events, because:
(a) The Ag-In-Cd RCCA/standard drive line weight continues to
meet the rod drop time of 2.7 seconds limit listed in Technical
Specification 3.1.5 (Rod Group Alignment Limits); and
(b) The reactivity difference was addressed for the impact on
core neutronics and safety analyses. It was determined that the
reactivity change can be accommodated within the bounds of the
current safety analysis limits using approved NRC methodology.
Future core designs will use an NRC approved methodology as the
means to demonstrate the continued safe operation of the plant with
the EP Ag-In-Cd RCCAs.
The change does not adversely affect the protective and
mitigative capabilities of the plant, nor does the change affect the
initiation or probability of occurrence of any accident. The SSCs
will continue to perform their intented 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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Watts Bar Unit 1 Technical Specification 4.2.2, Control Rod
Assemblies, is revised to include Ag-In-Cd material in addition to
the B4C control rod material. In addition to the absorber material
change, the replacement EP Ag-In-Cd RCCAs will be coupled with
Control Rod Drive Mechanism (CRDM) drive rod shafts which are
lighter than the CRDM drive rod shaft coupled to the B4C drive rod
shafts. Also, the EP Ag-In-Cd RCCAs are heavier than the B4C RCCAs
and have a different reactivity, or rod worth.
The EP Ag-In-Cd RCCAs are identical to the current RCCAs in
terms of form, fit, and function. The proposed changes will not
introduce any new failure mechanisms, malfunctions, or accident
initiators not already considered in the design and licensing basis.
The possibility of a new or different malfunction of safety-related
equipment is not created. No new accident scenarios, transient
precursors, or limiting single failures are introduced as a result
of these changes. There will be no adverse effects or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Watts Bar Unit 1 Technical Specification 4.2.2, Control. Rod
Assemblies, is revised to include Ag-In-Cd material in addition to
the B4C control rod material. In addition to the absorber material
change, the replacement EP Ag-In-Cd RCCAs will be coupled with
Control Rod Drive Mechanism (CRDM) drive rod shafts which are
lighter than the CRDM drive rod shaft coupled to the B4C drive rod
shafts. Also, the EP Ag-In-Cd RCCAs are heavier than the B4C RCCAs
and have a different reactivity, or rod worth. The changes in weight
and reactivity of the CRDM/RCCA on the design criteria and safety
analysis have been addressed.
The proposed changes regarding the Ag-In-Cd RCCAs do not involve
a significant reduction in a margin of safety, because:
(a) The Ag-In-Cd RCCA/standard drive line weight continues to
meet the rod drop time
[[Page 44027]]
of 2.7 seconds limit listed in Technical Specification 3.1.5 (Rod
Group Alignment Limits); and
(b) The reactivity difference was addressed for the impact on
core neutronics and safety analyses. It was determined that the
reactivity change can be accommodated within the bounds of the
current safety analysis limits using approved NRC methodology.
Future core designs will use an NRC approved methodology as the
means to demonstrate the continued safe operation of the plant with
the EP Ag-In-Cd RCCAs. Therefore, the proposed change does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Stephen J. Campbell.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management System (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334 Beaver
Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, Pennsylvania
Date of application for amendment: July 6, 2009, as supplemented on
March 10, 2010.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.6.3, ``Core Operating Limits Report,'' to allow
the use of the generically approved Topical Report, WCAP-16009-P-A,
``Realistic Large Break LOCA [Loss-of-Coolant Accident] Evaluation
Methodology Using Automated Statistical Treatment of Uncertainty
Method,'' for BVPS-1.
Date of issuance: July 1, 2010.
Effective date: As of the date of issuance, and shall be
implemented prior to startup following the fall 2010 maintenance and
refueling outage.
Amendment No: 286.
Facility Operating License No. DPR-66: The amendment revised the
License and TS.
Date of initial notice in Federal Register: December 1, 2009 (74 FR
62835). The March 8, 2010, supplement provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 1, 2010.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1 (NMP1), Oswego County, New York
Date of application for amendment: July 2, 2009.
Brief description of amendment: The amendment revises the TSs by
removing position indication for the relief valves and safety valves
from TS 3.6.11, ``Accident Monitoring Instrumentation.'' The amendment
would also correct an editorial error in the title of Table 4.6.11,
``Accident Monitoring Instrumentation Surveillance Requirement.''
Date of issuance: June 29, 2010.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 205.
Renewed Facility Operating License No. DPR-63: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: October 14, 2009 (74 FR
52826).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 2010.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of application for amendment: November 30, 2009.
Brief description of amendment: The amendment revises the emergency
diesel generator (DG) Completion Time for inoperable DGs in Technical
Specification (TS) 3.8.1, ``AC Sources Operating.'' The amendment
revises the Completion Time from 14 days to 72 hours for restoring one
or more inoperable DG(s) in one train to an operable status. The
amendment was requested because of the potential completion and startup
of the WBN Unit 2.
Date of issuance: July 6, 2010.
Effective date: As of the date of issuance and shall be implemented
after the issuance of the facility operating license for WBN Unit 2 and
prior to WBN Unit 2 entry into Mode 4, ``Hot Shutdown.''
Amendment No.: 84.
Facility Operating License No. NPF-90: Amendment revised the
License and TSs.
Date of initial notice in Federal Register: March 9, 2010 (75 FR
10830).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 6, 2010.
No significant hazards consideration comments received: No.
[[Page 44028]]
Dated at Rockville, Maryland, this 15th day of July 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-18078 Filed 7-26-10; 8:45 am]
BILLING CODE 7590-01-P