[Federal Register Volume 75, Number 75 (Tuesday, April 20, 2010)]
[Notices]
[Pages 20627-20644]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-8744]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0156]
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 25, 2010 to April 7, 2010. The last
biweekly notice was published on April 6, 2010 (75 FR 17439).
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
[[Page 20628]]
comment period or the notice period, it will publish in the Federal
Register a notice of issuance. Should the Commission make a final No
Significant Hazards Consideration Determination, any hearing will take
place after issuance. The Commission expects that the need to take this
action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules,
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of
Administrative Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be faxed to the RADB at 301-492-3446.
Documents may be examined, and/or copied for a fee, at the NRC's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area 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.
[[Page 20629]]
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through EIE, users will be required to install a Web
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHDProceeding/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: November 30, 2009.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.3.5, ``Engineered Safety Features
Actuation System Instrumentation,'' Table 3.3.5-1, to raise the
refueling water tank (RWT) low level allowable values for the
recirculation actuation signal (RAS); raise the minimum required RWT
volume shown in TS Figure 3.5.5-1; and implement a time-critical
operator action to close the RWT isolation valves, including
consideration of a potentially more limiting single failure of a low-
pressure safety injection pump to automatically stop, as designed, on
an RAS.
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 RWT is a passive component of the Chemical and Volume
Control System (CVCS) that supports ECCS [emergency core cooling
system] and CSS [containment spray system] operation to mitigate the
consequences of an accident. A[n] RAS is an active component of the
Engineered Safety Features Actuation System (ESFAS) that actuates
safety equipment to mitigate the consequences of a LOCA [loss-of-
coolant accident]. Neither of these components initiates an accident
previously evaluated. The RWT isolation valves are also components
of the CVCS; however, their closure was not previously credited for
RWT isolation following a[n] RAS. The proposed amendment will credit
closure of these valves following a[n] RAS to preclude the potential
for air entrainment in the ECCS and CS [containment spray] pump
suction piping for any LOCA scenario. The required isolation is
being performed as a time critical
[[Page 20630]]
operator action, which is consistent with ANSI/ANS-58.8-1984
[American National Standards Institute/American Nuclear Society
Standard 58.8-1984], Time Response Design Criteria for Safety-
Related Operator Actions, 1984 guidance. Although the change in the
closure requirement and the operator action could introduce
additional potential malfunctions, these malfunctions have been
evaluated and found not to initiate or have a significant adverse
affect on the mitigation or consequences of any accident previously
evaluated.
The proposed changes do not alter or prevent the ability of
structures, systems or components to perform their intended function
to mitigate the consequences of an initiating event within the
assumed acceptance limits. The proposed changes will ensure
continued performance of the ECCS and CS pumps following a LOCA by
precluding the potential for air entrainment in the pump suction
piping from the RWT after a[n] RAS.
The effect of the proposed changes to the RAS Allowable Values
and RWT minimum required level on the RWT structural design,
containment post-LOCA flood level, post-LOCA boron precipitation,
and containment sump pH remain within the limits assumed in the
design and accident analyses. The proposed license amendment does
not affect the source term, containment isolation, or radiological
release assumptions used in evaluating the radiological consequences
of an accident previously evaluated. Further, the proposed changes
do not increase the types or amounts of radioactive effluent that
may be released offsite. The proposed license amendment is
consistent with these analyses' assumptions and resultant
consequences.
The proposed amendment also recognizes and evaluates a different
single failure associated with the RWT drain down following a LOCA
than previously evaluated. It was determined this failure was of low
probability and did not adversely affect any previous bounding
analysis or the capability of the associated systems to perform
their design functions.
Therefore, the proposed license 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 license amendment does not involve or add any new
or different components to the plant and does not change any
accident initiators.
The proposed changes to the RAS Allowable Values and RWT minimum
required level will not change the design function of the RWT to
support ECCS and CSS operation following a LOCA. However, the
closure of the RWT isolation valves following a LOCA was not
previously credited. As a result, the credited RWT isolation valve
design function has been changed, and closure of these valves is now
credited to preclude the possibility of air entrainment in the ECCS
and CS pump suction piping for any LOCA scenarios. The credited
isolation is being performed as a time critical operator action,
which is consistent with ANSI/ANS 58.8 guidance. Although changes to
the valve closure requirement and the operator action introduce
additional potential malfunctions, these malfunctions have been
evaluated and found not to create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendment recognizes and evaluates a different
single failure associated with the RWT drain down following a LOCA
than previously evaluated. It was determined that this failure was
of low probability and did not adversely affect any previous
bounding analysis or create the possibility of a new or different
kind of accident from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed license amendment does not alter the manner in
which safety limits, limiting safety system settings, or limiting
conditions for operation are determined or implemented. The safety
analysis acceptance criteria are not affected by this amendment. The
proposed changes in the credited design function of the RWT
isolation valves, along with the change in the RAS Allowable Value
and RWT minimum required levels, continue to ensure sufficient RWT
water volume to enable the ECCS and CSS to satisfy required design
functions for all postulated LOCA break sizes. Therefore, these
changes do not impact the results of safety analyses.
The proposed changes to the RAS Allowable Values and minimum
required RWT level include appropriate instrument uncertainties and
are based on conservative analyses for establishing the required RWT
volumes. The proposed amendment will not result in plant operation
in a configuration outside of the design basis.
The proposed amendment recognizes and evaluates a different
single failure associated with the RWT drain down following a LOCA
than previously evaluated. It was determined this failure was of low
probability and did not adversely affect any previous bounding
analysis.
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
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.
Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendment request: January 29, 2010.
Description of amendment request: The amendment would modify the
existing Note within Technical Specification 3.4.10, ``Pressurizer
Safety Valves [PSVs],'' which covers operation in the applicable
portions of Mode 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No.
The proposed change, revising an existing NOTE within Technical
Specification 3.4.10 to allow the PSVs lift settings to be outside
LCO [Limiting Condition for Operation] values, as a result of
temperature related drift, while the Unit is in applicable portions
of Mode 3 for periods up to 36 hours, does not change the design
function or operation of the PSVs and it does not change the way the
PSVs are maintained, tested, or inspected. In addition the proposed
change does not change any of the evaluated accidents in our Updated
Final Safety Analysis Report, does not change PSV lift settings, or
impact the ability of the PSVs to perform their safety function
during evaluated accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No.
The proposed change, revising an existing NOTE within Technical
Specification 3.4.10 to allow the PSVs lift settings to be outside
LCO values, as a result of temperature related drift, while the Unit
is in applicable portions of Mode 3 for periods up to 36 hours, does
not change the PSVs design function to maintain RCS [reactor coolant
system] pressure below the RCS pressure Safety Limit of 2750 psia
during design basis accidents nor does it affect the PSVs ability to
perform this design function. The proposed change does not require
any modification to the plant or change equipment operation or
testing. It also does not create any credible new failure
mechanisms, malfunctions, or accident initiators that would cause an
accident not previously considered.
Therefore the proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
[[Page 20631]]
No.
The proposed change, revising an existing NOTE within Technical
Specification 3.4.10 to allow the PSVs lift settings to be outside
LCO values, as a result of temperature related drift, while the Unit
is in applicable portions of Mode 3 for periods up to 36 hours, does
not involve a significant reduction in the margin of safety in
maintaining RCS pressure below Safety Limits of 2750 psia during
design basis accidents. The analysis conducted in support of this
proposed change evaluated the ability of the PSVs to maintain an
adequate safety margin when required in applicable Mode 3 conditions
despite the identified temperature related lift setting drift. The
analysis identified that there were no credible design accident
scenarios, when in the applicable Mode 3 conditions, that challenged
the PSVs to respond in order to maintain an adequate safety margin
to the reactor coolant Safety Limit of 2750 psia.
Therefore the proposed change does not involve a significant
reduction in the margin of safety of maintaining RCS pressure below
the RCS pressure Safety Limit.
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
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Nancy L. Salgado.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: January 4, 2010.
Description of amendment request: The proposed amendment would
revise the Core Spray flow requirement in Technical Specifications
Surveillance Requirements 3.5.1.8 and 3.5.2.6 from 6,350 to 5,725
gallons per minute consistent with the flow assumed in the Emergency
Core Cooling System (ECCS) safety analyses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The minimum performance requirements of the low pressure
Emergency Core Cooling System (ECCS) pumps, including the Core Spray
pumps, are determined through application of the 10 CFR 50, Appendix
K methodology to ensure the criteria of 10 CFR 50.46 are satisfied.
The surveillance testing of the Core Spray pumps is performed
periodically in accordance with the ASME Code, Section XI verifies
that two Core Spray pumps in parallel operation within a single
division develop sufficient discharge pressure at the Technical
Specification required flow to overcome the elevation head pressure
between the pump suction and the vessel discharge, the piping
friction losses, and TS SR specified Reactor Pressure Vessel
pressure. The acceptance criteria necessary to satisfy the revised
TS SRs would be established in the plant design basis in the form of
the minimum required pump performance defined for a range of flow
about the specified TS SR flow. Detroit Edison intends to continue
TS SR and IST pump testing at the current IST pump baseline flow and
establish compliance with the TS SR by comparing the measured
performance against the design minimum pump curve. In this manner,
the minimum actual delivered divisional Core Spray pump performance
is assured to meet or exceed that required by the Appendix K safety
analyses. These performance requirements are unchanged and are met
by the proposed change.
The bases for the core spray flow requirements in the Technical
Specifications Surveillance Requirements are unchanged. The
requirements are selected based on the flow values assumed and used
in the current ECCS safety analyses. The value proposed for core
spray divisional (2 pump) flow is consistent with the inputs used
for ECCS safety analyses performed for the current licensed power
level.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change revises the Technical Specification
Surveillance Requirements for Core Spray flow to be consistent with
the accident analysis. No physical changes are being made to the
installed core spray system. The proposed surveillance requirements
are consistent with those used in the accident analyses which
analyze the effect of Core Spray system performance for the accident
conditions for which the system is designed to respond. No new or
different accident scenarios are created by this change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The Core Spray system has historically been capable of meeting
the Core Spray Technical Specification Surveillance Requirements.
However, correction of non-conservative errors in the system
hydraulic calculation and the identification of a non-conservative
bias in the test flow instrument calibration have eroded the test
margin such that it is possible that the Technical Specification
Surveillance Requirements may not be satisfied for some
surveillances and at the same time maintain a relatively large
margin compared to the minimum performance assumed in the ECCS
safety analyses. These non-conservative errors or biases have always
existed, but have not always been specifically accounted for in the
surveillance testing acceptance criteria. Since there is no change
in the Technical Specification bases associated with the requested
change, there is no real change in the margin provided in the system
design or analyses. The proposed change makes the margin between the
current Core Spray Technical Specification Surveillance Requirements
and the performance assumed in the plant safety analyses available
as a design and test margin. The minimum required performance
necessary to satisfy the Core Spray Technical Specification
Surveillance Requirements will be established in the plant design
basis with the minimum required pump performance adjusted upward as
necessary to account for instrument uncertainty and bias as well as
differences between assumed accident and actual test operating
conditions.
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: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: November 23, 2009, as supplemented by
letter dated March 18, 2010.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TS) requirements for testing of
the James A. FitzPatrick Nuclear Power Plant (JAFNPP) Safety/Relief
Valves (SRVs) by replacing the current requirement to manually actuate
each SRV during plant startup with a requirement to verify that each
valve is capable of being opened. The proposed amendment would change
both TS Surveillance Requirements (SRs) 3.4.3.2 and 3.5.1.13 to verify
that each required valve ``is capable of being opened.'' The current
Frequency for both TS SRs is ``24 months on a STAGGERED TEST BASIS for
each valve solenoid''; this
[[Page 20632]]
would be changed to state, ``In accordance with the Inservice Testing
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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed change does not modify the method of demonstrating
the Operability of the Safety/Relief Valves (SRVs) in both the
safety and relief modes of operation. As currently stated in the
Bases ``...valve OPERABILITY and the setpoints for overpressure
protection are verified, per ASME Code requirements, prior to valve
installation.'' The proposed change does modify the method for
demonstrating the proper mechanical functioning of the SRVs and that
the valves and discharge lines are free of obstructions. The SRVs
are required to function in the safety mode to prevent
overpressurization of the reactor vessel and reactor coolant system
pressure boundary during various analyzed transients, including Main
Steam Isolation Valve closure. SRVs associated with the Automatic
Depressurization System are also required to function in the relief
mode to reduce reactor pressure to permit injection by low pressure
Emergency Core Cooling System (ECCS) pumps during certain reactor
coolant pipe break accidents. The current testing method
demonstrates the proper mechanical functioning of the SRVs in both
modes through manual actuation of the SRVs. The proposed new testing
method demonstrates both Operability and proper mechanical
functioning using a series of overlapping tests that demonstrate
proper functioning of the SRV stages and supporting control
components. This proposed testing method results in acceptable
demonstration of the SRV functions in both the safety and relief
modes, and therefore provides assurance that the probability of SRV
failure will not increase. None of the accident safety analyses is
affected by the requested Technical Specifications (TS) changes.
Therefore, the consequences of accidents mitigated by the SRVs will
not increase.
Certain SRV malfunctions are included in the FSAR [final safety
analysis report] safety analyses. Specifically, the plant safety
analyses include the inadvertent opening of an SRV and a stuck open
SRV. By not actuating the SRVs during plant operation for testing
and thus reducing the incidence of pilot stage leakage of the SRVs,
the proposed testing eliminates a contributor to these events.
Based on these considerations, the proposed test method does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed change modifies the method of testing of the SRVs,
but does not alter the functions or functional capabilities of the
SRVs. Testing under the proposed method is performed in offsite test
facilities or in the plant during outage periods when the SRV
functions are not required. Existing analyses address events
involving an SRV inadvertently opening or failing to reclose.
Analyses also address the likelihood and consequences of failure of
one or more SRVs to open. The proposed change does not introduce any
new failure mode, and therefore, does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
Overpressure protection of the reactor coolant pressure boundary
is based on the SRV setpoints and total relief capacity. Setpoint is
verified at an offsite testing facility; this requirement is not
altered by the proposed change. Relief capacity of each SRV is
determined by valve geometry, which is also not altered by the test
methods. The margin of safety in the Loss of Coolant Accident
analysis due to operation of the Automatic Depressurization System
is also based on total relief capacity of the associated SRVs. The
proposed change in surveillance test methods demonstrates the
operability of the SRVs, but does not alter the critical parameters
that affect the margin of safety in analyses involving the SRV
functions. Therefore, the proposed change does not involve a
significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy L. Salgado.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 22, 2010.
Description of amendment request: The proposed amendment will allow
implementation of leak-before-break (LBB) on the Waterford Steam
Electric Station, Unit 3 (Waterford 3) pressurizer surge line. The
licensee will be replacing the two Waterford 3 steam generators (SGs)
during the forthcoming spring 2011 refueling outage. Based on design
changes in the replacement SGs, piping systems will require rerouting
in the SG cavity area. Due to the existing dynamic piping protection
associated with the pressurizer surge line, rerouting of the
replacement SG blowdown line cannot be effectively performed without
the elimination of dynamic protection for the pressurizer surge line.
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 uses an approved leak-before-break (LBB)
fracture mechanics methodology, in accordance with 10CFR50 [Title 10
of the Code of Federal Regulations, Part 50], Appendix A, General
Design Criterion (GDC) 4 to demonstrate that the probability of
fluid system rupture for these lines attached to the Reactor Coolant
System (RCS) is extremely low under conditions associated with the
design basis for the piping. The proposed change does not adversely
affect accident initiators or precursors nor significantly alter the
design assumptions, conditions, and configuration of the facility or
the manner in which the plant is operated and maintained. Overall
protection system performance will remain within the bounds of the
previously performed accident analyses. The design of the protection
systems will be unaffected. The Reactor Protection System (RPS) and
Emergency Core Cooling System (ECCS) will continue to function in a
manner consistent with the plant design basis. All design, material,
and construction standards that were applicable prior to the request
are maintained. There will be no change to normal plant operating
parameters or accident mitigation performance. The proposed
amendment will not alter any assumptions or change any mitigation
actions in the radiological consequence evaluations in the FSAR
[Final Safety Analysis Report].
Therefore, this 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 create the possibility of a new or
different kind of accident, since it provides an NRC acceptable
alternate means for demonstrating that the probability of a fluid
system rupture is extremely small. There are no changes in the
methods by which any safety-related plant
[[Page 20633]]
system performs its safety function. No new accident scenarios,
transient precursors, failure mechanisms, or limiting single
failures are introduced as a result of this amendment. There will be
no adverse effect or challenges imposed on any safety-related system
as a result of this amendment. LBB methodology per GDC-4 still
requires that ECCS, containment, and equipment qualification (EQ)
requirements be maintained consistent with the original postulated
accident assumptions. Only protection from dynamic effects is
modified.
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 changes apply conservative approved analytical
methods to demonstrate that the probability of a fluid system
rupture is very low. This analysis retains substantial margins to
assure that pipe rupture is extremely low and justifies differences
in protection from dynamic effects with these extremely low
probability ruptures. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions. For overall ECCS,
containment, and EQ requirements, there will be no changes to the
assumed margins.
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
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 22, 2010.
Description of amendment request: The proposed amendment would add
valve SI-4052A (Reactor Coolant Loop (RCL) 2 Shutdown Cooling (SDC)
suction inside containment bypass isolation) and valve SI-4052B (RCL 1
SDC suction inside containment bypass isolation) to Technical
Specification (TS) Table 3.4-1, ``Reactor Coolant System Pressure
Isolation Valves.'' The purpose of this line is to equalize the SDC
system pressure down stream of valve SI-405A (RCL 2 SDC suction inside
containment isolation) and valve SI-405B (RCL 1 SDC suction inside
containment isolation) in order to minimize the pressure transient in
the system when valves SI-405A(B) are opened.
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 addition of the bypass fill line will decrease the
likelihood of a pressure transient in the Shutdown Cooling System
suction piping which increases the reliability of the Shutdown
Cooling System. Once this change is installed valves SI-405A(B) and
SI-4052A(B) become parallel inside containment isolation valves in
the shutdown cooling system suction lines. The configuration of SI-
405A(B) and SI-4052A(B) includes interlocks such that these valves
cannot be inadvertently opened with the RCS [reactor coolant system]
above the design pressure of the shutdown cooling system. This
change does not affect the capability of these valves to isolate the
RCS from SDC. Therefore, there is no credible mechanism by which
this change can introduce an inter-system LOCA [loss-of-coolant
accident] (ISLOCA) different than previously evaluated in the UFSAR
[Updated Final Safety Analysis Report]. These features are,
discussed in FSAR [Final Safety Analysis Report] section 7.6.1.1.2.
Therefore, this 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.
Once this change is installed valves SI-405A(B) and SI-4052A(B)
become parallel inside containment isolation valves in the shutdown
cooling system suction lines. SI-4052A(B) and its associated lines
and valves are designed to the same requirements as SI-405A(B) and
its associated lines. The previously evaluated SI-405A(B) failure
modes bound those failure modes possible by SI-4052A(B). Thus, no
failure of SI-4052A(B) exists that would be different or more severe
than SI-405A(B),
This proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment adds SI-4052A(B) to Technical
Specification Table 3.4-1. The change also adds an allowed leakage
limit to SI-4052A(B) consistent with NUREG-1432 guidance.
Since the SI-4052A(B) leakage limit is commensurate with the
valve size, this does not represent 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 Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 22, 2010.
Description of amendment request: Entergy Operations, Inc. (the
licensee), will be replacing the two Waterford Steam Electric Station,
Unit 3 (Waterford 3) steam generators (SGs) during the 17th refueling
outage which will commence in the spring of 2011. The existing
Waterford 3 SG program under Technical Specification (TS) 6.5.9
contains an alternate repair criterion for SG tube inspections that is
no longer applicable to the replacement SGs. The proposed amendment
will modify TS 6.5.9, ``Steam Generator (SG) Program,'' and TS 6.9.1.5,
``Steam Generator Tube Inspection Report,'' to eliminate currently
allowed SG tube alternate repair criteria and to modify the SG tube
inservice inspection frequency.
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 continues to implement the Waterford 3 Steam
Generator Program performance criteria for tube structural
integrity, accident induced leakage, and operational leakage for the
replacement SGs. Meeting the performance criteria provides
reasonable assurance that the replacement SG tubing will remain
capable of fulfilling its specific safety function of maintaining
reactor coolant system (RCS) pressure boundary integrity throughout
each operating cycle and in the unlikely event of a design basis
accident.
[[Page 20634]]
The Steam Generator Tube Rupture (SGTR) is the primary accident
analysis associated with SG tube integrity. The replacement SG
tubing contains improved materials that will reduce the likelihood
of tubing flaws. The proposed change to remove alternate repair
criteria from the SG inspection program does not affect the design
of the replacement SGs, their method of operation, operational
leakage limits, or primary coolant chemistry controls. Therefore,
the proposed change does not affect the probability of a SGTR
accident. The SGs will be designed with substantial margin to burst.
The SG tube inspection repair limit will also identify potential
flaws before they become a safety concern. The extension of the SG
tube inspection frequency after initial inspection is based on the
low likelihood of having potential tube flaws and is considered to
be an acceptable inspection period to preserve pressure boundary
integrity. As a result, there will be no affect on the previous dose
analysis reported in the FSAR [Final Safety Analysis Report] and the
consequences of any accident are unchanged.
Therefore, this 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.
Steam generator tube rupture events have already been postulated
and analyzed in the Waterford 3 FSAR. The proposed change does not
affect the design of the SGs, their method of operation, or primary
or secondary coolant chemistry controls. Additionally, the proposed
amendment does not impact any other plant systems or components. The
TSs have established SG tube inspection requirements which assure
that potential tubing flaws will be detected prior to affecting tube
integrity and the RCS pressure boundary. Therefore, the proposed
change does not create the possibility of a new or different type of
accident from any accident previously evaluated.
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 structural integrity, accident induced leakage, and
operational leakage performance criteria required by the Waterford 3
TSs provide substantial design margin for assuring SG tube integrity
against the possibility of a SG tube pressure boundary failure. The
proposed change removes an existing alternate repair criterion that
is not applicable to the replacement SGs and establishes appropriate
SG tube subsequent inspection periods consistent with the new SG
tubing design. The replacement SGs will continue to meet their
required performance criteria. The Waterford 3 SG tube inspection
program will assure that this margin is maintained through the
operational life of the plant.
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
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Date of amendment request: February 15, 2010.
Description of amendment request: This amendment request involves
the adoption of Nuclear Regulatory Commission (NRC)-approved changes to
the Standard Technical Specifications (STS) for Westinghouse plants
(NUREG-1431), to allow relocation of specific TS surveillance
frequencies to a licensee-controlled program. The proposed changes are
described in Technical Specification Task Force (TSTF) Traveler, TSTF-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee
Control--Risk Informed Technical Specification Task Force (RITSTF)
Initiative 5b,'' as announced in the Notice of Availability published
in the Federal Register on July 6, 2009 (74 FR 31996). Additionally,
the proposed changes would add a new program, the Surveillance
Frequency Control Program, to TS Section 5, Administrative Controls.
The changes are applicable to licensees using the probabilistic risk
guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-
10, Revision 1, ``Risk-Informed Technical Specifications Initiative 5b,
Risk-Informed Method for Control of Surveillance Frequencies.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration adopted by the licensee is
presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
significantly increased. The systems and components required by the
Technical Specifications for which the surveillance frequencies are
relocated are still required to be operable, meet the acceptance
criteria for the surveillance requirements, and be capable of
performing any mitigation function assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated
are not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the Updated Final Safety Analysis Report and Bases
to the Technical Specifications), because these are not affected by
changes to the surveillance frequencies. Similarly, there is no
impact to safety analysis acceptance criteria as described in the
plant-licensing basis. To evaluate a change in the relocated
surveillance frequency, EGC will perform a probabilistic risk
evaluation using the guidance contained in NRC approved NEI 04-10,
Revision 1 in accordance with the TS Surveillance Frequency Control
Program. NEI 04-10, Revision 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase
of proposed changes to surveillance frequencies consistent with
Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the three standards of 10
CFR 50.92(c) are
[[Page 20635]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: February 15, 2010.
Description of amendment request: This amendment request involves
the adoption of Nuclear Regulatory Commission (NRC)-approved changes to
the Standard Technical Specifications (STS) for Westinghouse plants
(NUREG-1431), to allow relocation of specific TS surveillance
frequencies to a licensee-controlled program. The proposed changes are
described in Technical Specification Task Force (TSTF) Traveler, TSTF-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee
Control--Risk Informed Technical Specification Task Force (RITSTF)
Initiative 5b,'' as announced in the Notice of Availability published
in the Federal Register on July 6, 2009 (74 FR 31996). Additionally,
the proposed changes would add a new program, the Surveillance
Frequency Control Program, to TS Section 5, Administrative Controls.
The changes are applicable to licensees using the probabilistic risk
guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-
10, Revision 1, ``Risk-Informed Technical Specifications Initiative 5b,
Risk-Informed Method for Control of Surveillance Frequencies.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration adopted by the licensee is
presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
significantly increased. The systems and components required by the
Technical Specifications for which the surveillance frequencies are
relocated are still required to be operable, meet the acceptance
criteria for the surveillance requirements, and be capable of
performing any mitigation function assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated
are not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the Updated Final Safety Analysis Report and Bases
to the Technical Specifications), because these are not affected by
changes to the surveillance frequencies. Similarly, there is no
impact to safety analysis acceptance criteria as described in the
plant-licensing basis. To evaluate a change in the relocated
surveillance frequency, EGC will perform a probabilistic risk
evaluation using the guidance contained in NRC approved NEI 04-10,
Revision 1 in accordance with the TS Surveillance Frequency Control
Program. NEI 04-10, Revision 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase
of proposed changes to surveillance frequencies consistent with
Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
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. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Date of amendment request: February 4, 2010.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.3.61, ``Primary Containment
Isolation Instrumentation,'' Table 3.3.6.1-1, ``Primary Containment
Isolation Instrumentation,'' Function 6.a, ``Shutdown Cooling System
Isolation, Recirculation Line Water Temperature--High,'' to enable
implementation of a modification that replaces the temperature-based
isolation instrumentation with reactor pressure-based isolation
instrumentation. The proposed modification will address instrumentation
reliability problems that have led to interruptions of Shutdown Cooling
(SDC) system operation, leading to unplanned heat-up of reactor coolant
while the reactor was in operational Modes 3 and 4.
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 license amendment implements a revised process
parameter and the associated Allowable Value (AV) for the DNPS Units
2 and 3 SDC system isolation function 6.a in TS Table 3.3.6.1-1.
The proposed changes to the isolation function do not affect the
probability of any event initiators at the facilities. This
isolation function is provided for equipment protection to prevent
exceeding the system design temperature. The isolation function is
not credited or assumed in the accident or transient analysis in the
Updated Final Safety Analysis Report (UFSAR).
The proposed changes will not degrade the performance of, or
increase the number of challenges imposed on, safety-related
equipment that is assumed to function during an accident situation.
The SDC system and the isolation function that is being revised are
not safety related and are not credited to function during an
accident situation. The proposed changes will not alter any
assumptions or change any mitigation actions in the radiological
consequence evaluations in the UFSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 20636]]
(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 license amendment implements a revised process
parameter and AV for the DNPS Units 2 and 3 SDC system isolation
function 6.a in TS Table 3.3.6.1-1. The proposed change enables
implementation of a modification that will enhance the reliability
of instrumentation used to protect the functionality and integrity
of the non safety-related SDC system. There is no alteration to the
parameters within which the plant is normally operated or in the
setpoints that initiate protective or mitigative actions. As a
result, no new failure modes are being introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The proposed license amendment revises a process parameter and
AV for the DNPS Units 2 and 3 SDC system isolation function 6.a in
TS Table 3.3.6.1-1.
The margin of safety is established through the design of the
plant structures, systems, and components (SSCs), the parameters
within which the plant is operated, and the setpoints for the
actuation of equipment relied upon to respond to an accident.
The proposed change to the SDC system isolation instrumentation
function for the SDC system does not change the SSCs, operational
parameters, or actuation setpoints for equipment that is relied upon
to respond to an accident. Both the SDC system and the isolation
function that is being revised are non-safety related and are not
credited to function during an accident situation.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Date of amendment request: February 16, 2010.
Description of amendment request: The proposed amendments would
modify the DNPS Units 2 and 3, Technical Specifications (TS) by
relocating specific surveillance frequencies to a licensee-controlled
program with the adoption of Technical Specification Task Force (TSTF)-
425, ``Relocate Surveillance Frequencies to Licensee Control--Risk
Informed Technical Specification Task Force (RITSTF) Initiative 5b,''
Revision 3. Additionally, the change would add a new program, the
``Surveillance Frequency Control Program [SFCP],'' to TS Section 5,
``Administrative Controls.''
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. The licensee reviewed the proposed No Significant
Hazards Consideration (NSHC) determination published in the Federal
Register dated July 6, 2009 (74 FR 31996).
The licensee has concluded that the proposed NSHC presented in the
Federal Register notice is applicable to DNPS, Units 2 and 3. The
proposed NSHC is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements (SRs) to licensee control under a
new SFCP. Surveillance frequencies are not an initiator to any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased. The
systems and components required by the TS for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the SRs, and be capable of performing
any mitigation function assumed in the accident analysis. As a
result, the consequences of any accident previously evaluated are
not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to the
TS), because these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, EGC will
utilize the guidance contained in NRC-approved NEI 04-10, in
accordance with the TS SFCP. NEI 04-10, Revision 1 methodology
provides reasonable acceptance guidelines and methods for evaluating
the risk increase of proposed changes to surveillance frequencies
consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: February 15, 2010.
Description of amendment request: The proposed amendments would
modify the LaSalle County Station (LSCS) Technical Specifications (TS)
by relocating specific surveillance frequencies to a licensee-
controlled program with the implementation of Nuclear Energy Institute
(NEI) 04-10.
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:
1. Do the proposed changes involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result,
the probability of any
[[Page 20637]]
accident previously evaluated is not significantly increased. The
systems and components required by the Technical Specifications for
which the surveillance frequencies are relocated are still required
to be operable, meet the acceptance criteria for the surveillance
requirements, and be capable of performing any mitigation function
assumed in the accident analysis. As a result, the consequences of
any accident previously evaluated are not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the Updated Final Safety Analysis Report and Bases
to the Technical Specifications), because these are not affected by
changes to the surveillance frequencies. Similarly, there is no
impact to safety analysis acceptance criteria as described in the
plant licensing basis. To evaluate a change in the relocated
surveillance frequency, EGC will perform a probabilistic risk
evaluation using the guidance contained in NRC approved NEI 04-10,
Revision 1 in accordance with the TS Surveillance Frequency Control
Program. NEI 04-10, Revision 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase
of proposed changes to surveillance frequencies consistent with
Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: February 22, 2010.
Description of amendment request: The proposed amendments would
revise Technical Specification 3.1.7, ``Standby Liquid Control (SLC)
System,'' to extend the completion time associated with Condition B
from 8 hours to 72 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment revises Technical Specification (TS)
3.1.7, ``Standby Liquid Control (SLC) System,'' to extend the
completion time (CT) associated with Condition B (i.e., ``Two SLC
subsystems inoperable.'') from eight hours to 72 hours.
The proposed change is based on a risk-informed evaluation
performed in accordance with Regulatory Guides (RG) 1.174, ``An
Approach for Using Probabilistic Risk Assessment in Risk-Informed
Decisions On Plant-Specific Changes to the Licensing Basis,'' and RG
1.177, ``An Approach for Plant-Specific, Risk-Informed Decision-
making: Technical Specifications.''
The proposed amendment modifies an existing CT for a dual-train
SLC system inoperability. The condition evaluated, the action
requirements, and the associated CT do not impact any initiating
conditions for any accident previously evaluated.
The proposed amendment does not increase postulated frequencies
or the analyzed consequences of an Anticipated Transient Without
Scram (ATWS). Requirements associated with 10 CFR 50.62 will
continue to be met. In addition, the proposed amendment does not
increase postulated frequencies or the analyzed consequences of a
large-break loss-of-coolant accident for which the SLC system will
be used for pH control (i.e., upon NRC approval of an August 26,
2008 proposed LSCS license amendment regarding the adoption of an
alternate source term methodology). The extended CT provides
additional time to implement actions in response to a dual-train SLC
system inoperability, while also minimizing the risk associated with
continued operation. Therefore, the proposed change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment revises TS 3.1.7 to extend the CT
associated with Condition B from eight hours to 72 hours. The
proposed amendment does not involve any change to plant equipment or
system design functions. This proposed TS amendment does not change
the design function of the SLC system and does not affect the
system's ability to perform its design function. The SLC system
provides a method to bring the reactor, at any time in a fuel cycle,
from full power and minimum control rod inventory to a subcritical
condition with the reactor in the most reactive xenon free state
without taking credit for control rod movement. Required actions and
surveillance requirements are sufficient to ensure that the SLC
system functions are maintained. No new accident initiators are
introduced by this amendment. Therefore, the proposed amendment does
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment revises TS 3.1.7 to extend the CT
associated with Condition B from eight hours to 72 hours. The
proposed amendment does not involve any change to plant equipment or
system design functions. The margin of safety is established through
the design of the plant structures, systems, and components, the
parameters within which the plant is operated, and the setpoints for
the actuation of equipment relied upon to respond to an event.
Safety margins applicable to the SLC system include pump
capacity, boron concentration, boron enrichment, and system response
timing. The proposed amendment does not modify these safety margins
or the point at which SLC is manually initiated, nor does it affect
the system's ability to perform its design function. In addition,
the proposed change complies with the intent of the defense-in-depth
philosophy and the principle that sufficient safety margins are
maintained, consistent with RG 1.177 requirements (i.e., Section C,
``Regulatory Position,'' paragraph 2.2, ``Traditional Engineering
Considerations'').
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
[[Page 20638]]
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station (QCNPS), Units 1 and 2, Rock Island
County, Illinois
Date of amendment request: February 16, 2010.
Description of amendment request: The proposed amendments would
modify the QCNPS Units 1 and 2, Technical Specifications (TS) by
relocating specific surveillance frequencies to a licensee-controlled
program with the adoption of Technical Specification Task Force (TSTF)-
425, ``Relocate Surveillance Frequencies to Licensee Control--Risk
Informed Technical Specification Task Force (RITSTF) Initiative 5b,''
Revision 3. Additionally, the change would add a new program, the
``Surveillance Frequency Control Program [SFCP],'' to TS Section 5,
``Administrative Controls.''
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. The licensee reviewed the proposed No Significant
Hazards Consideration (NSHC) determination published in the Federal
Register dated July 6, 2009 (74 FR 31996).
The licensee has concluded that the proposed NSHC presented in the
Federal Register notice is applicable to QCNPS, Units 1 and 2. The
proposed NSHC is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements (SRs) to licensee control under a
new SFCP. Surveillance frequencies are not an initiator to any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased. The
systems and components required by the TS for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the SRs, and be capable of performing
any mitigation function assumed in the accident analysis. As a
result, the consequences of any accident previously evaluated are
not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to the
TS), because these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, EGC will
utilize the guidance contained in NRC-approved NEI 04-10, in
accordance with the TS SFCP. NEI 04-10, Revision 1 methodology
provides reasonable acceptance guidelines and methods for evaluating
the risk increase of proposed changes to surveillance frequencies
consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: December 14, 2009.
Description of amendment request: The proposed amendment would
remove the structural integrity requirements contained in Technical
Specifications (TSs) 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) and their
associated Bases; incorporate changes to accident monitoring
instrumentation for consistency with NUREG-1432 actions and allowed
outage times for conditions that drive a unit to hot shutdown; and
administrative corrections based on obvious typos, previous amendments,
or obsolete requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change to remove structural integrity controls
from the TSs does not impact any mitigation equipment or the ability
of the RCS [reactor coolant system] pressure boundary to fulfill any
required safety function. The proposed change will continue to
ensure the requirements of 10 CFR 50.55a are maintained as specified
in TS 4.0.5 and the new administrative TS program for RCP [reactor
coolant pump] flywheel inspections. The changes to the accident
instrumentation actions and allowed outage time have no appreciable
effect on accident initiation or mitigation. Since no other accident
mitigation or initiators are impacted by this change, no design
basis accidents are affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
The proposed change will not alter the plant configuration or
change the manner in which the plant is operated. Structural
integrity will continue to be maintained as required by 10 CFR
50.55a and specified in TS 4.0.5 and the new administrative TS
program for RCP flywheel inspections. Accident monitoring
instrumentation does not contribute to failure modes. No new failure
modes are being introduced by the proposed change.
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
the margin of safety?
Removing TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) from the
TSs does not reduce the controls that are required to maintain the
structural integrity of ASME Code Class 1, 2, or 3 components. There
is no increase with any accident mitigation risk associated with the
accident monitoring instrumentation TS changes as the proposed
allowed outage times and the intervening step through HOT STANDBY
are consistent with the equivalent to NUREG-1432 completion times
and actions for post accident instrumentation and are equal to or
more conservative than the current TS requirements. No other safety
margins are impacted due to the proposed change.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
[[Page 20639]]
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: M.S. Ross, Attorney, Florida Power &
Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Acting Branch Chief: Douglas A. Broaddus.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: February 25, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirement (SR)
3.8.1.9, Diesel Generator (DG) Load Test, to correct a non-
conservative power factor (PF) value and to add a new note
consistent with TS Task Force (TSTF) traveler TSTF-276-A, Revision
2, ``Revise DG Full Load Rejection Test.''
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Performing a surveillance that tests the DG is not a precursor
of any accident previously evaluated. Revising the PF limit to be
more conservative, and relaxing the requirement to maintain PF when
paralleled to offsite power does not significantly affect the method
of performing the surveillances such that the probability of an
accident would be affected. These changes only affect surveillances
of mitigative equipment and, therefore, do not have an impact on the
probability of an accident previously evaluated.
Revising the surveillances by specifying a more conservative PF
value ensures the DG's will provide the power assumed in
calculations of design basis accident mitigation. Relaxing the
requirement to maintain PF when paralleled to offsite power does not
affect performance of the DG under accident conditions. The
performance of the surveillances ensures that mitigative equipment
is capable of performing its intended function, and therefore, the
change does not involve a significant increase in the consequences
of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes 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 limiting
single failures are introduced as a result of the proposed changes.
The systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on a safety-related system or component and do not challenge
the performance or integrity of safety related systems. As such, it
does not introduce a mechanism for initiating a new or different
accident than those described in the USAR [updated safety analysis
report].
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes will continue to ensure the DGs are able to
perform their design function as assumed in calculations that
evaluate their function during design basis accidents. Decreasing
the PF limit for testing will not affect the design or functioning
of the DGs. The increased reactive loading required to maintain the
PF below the limit is small and well within DG capability. Based on
this, the ability of CNS [Cooper Nuclear Station] to mitigate the
design basis accidents that rely on operation of the DG's is not
adversely impacted. Revising the PF increases the margin of safety
by specifying a more conservative value for the PF limit. Therefore,
NPPD [Nebraska Public Power District] concludes these 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: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
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].
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: July 13, 2007, as supplemented
by letters dated. July 13, 2007, September 30, 2008, March 5, 2009,
March 23, 2009, March 1, 2010, and March 5, 2010.
Brief description of amendment: The license amendment revises the
Millstone Power Station, Unit No. 3 (MPS3) spent fuel pool (SFP)
storage requirements. The July 13, 2007, license amendment request
proposed a stretch power uprate (SPU) of MPS3. Included in a supplement
dated July 13, 2007, was a request to amend the MPS3 SFP storage
requirements. The July 13, 2007, request was noticed in the Federal
Register on January 15, 2008 (73 FR 2549). By letter dated March 5,
2008, Dominion Nuclear Connecticut, Inc. (DNC) separated the MPS3 SFP
storage
[[Page 20640]]
requirements request from the MPS3 SPU request. The request to revise
the MPS3 SFP storage requirements was re-noticed on September 8, 2009
(74 FR 46241) using the original significant hazards consideration,
specific to the request to revise the SFP storage.
Date of issuance: March 26, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 248.
Renewed Facility Operating License No. NPF-49: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: January 15, 2008 (73 FR
2549) and September 8, 2009 (74 FR 46241). The supplemental letters
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination as
published in the Federal Register (73 FR 2549). The SFP LAR no
significant hazards consideration determination was noticed a second
time, separate from the MPS3 SPU (74 FR 46241).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2010.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana
Parish, Louisiana
Date of amendment request: June 29, 2010.
Brief description of amendment: The amendment revised the RBS
Technical Specification (TS) 5.5.6, ``Inservice Testing Program.'' TS
5.5.6 contains references to the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code, Section XI as the
source for the inservice testing (IST) of ASME Code Class 1, 2, and 3
pumps and valves. The proposed changes delete the references to Section
XI of the ASME Code and incorporate references to the ASME Code for
Operation and Maintenance of Nuclear Power Plants (OM Code). In
addition, the amendment changes will limit applying Surveillance
Requirement (SR) 3.0.2 to surveillances with a frequency of 2 years or
less. These changes are consistent with the changes identified in the
Improved Standard Technical Specifications (ISTS) in Technical
Specification Task Force Traveler (TSTF) Change Travelers TSTF-479,
``Changes to Reflect Revision of 10 CFR 50.55a,'' and TSTF-497, ``Limit
Inservice Testing Program 3.0.2 Application to Frequencies of 2 Years
or Less.''
Date of issuance: March 31, 2010.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 167.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 25, 2009 (74 FR
42928).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 2010.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station (GGNS), Unit 1, Claiborne
County, Mississippi
Date of application for amendment: October 27, 2009.
Brief description of amendment: The amendment revised Technical
Specification (TS) Section 2.1.1, ``Reactor Core SLs [Safety Limits],''
Subsection 2.1.1.2, to change the two recirculation loop safety limit
for minimum critical power ratio (SLMCPR) from 1.08 to 1.09 and the
single recirculation loop SLMCPR from 1.10 to 1.12. The changes to the
TSs are necessary as a result of the GGNS Cycle 18 cycle-specific
SLMCPR calculations.
Date of issuance: March 25, 2010.
Effective date: As of the date of issuance and shall be implemented
after the current cycle (Cycle 17) is completed and prior to the
operation of Cycle 18.
Amendment No: 184.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 5, 2010 (75 FR
461).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 25, 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. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois,
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station
(QCNPS), Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: April 7, 2009, as supplemented
by letter dated October 5, 2009.
Brief description of amendments: The amendments delete a footnote
from DNPS Technical Specification (TS) 3.4.5, ``RCS Leakage Detection
Instrumentation,'' that was incorporated as part of a limited duration
emergency license amendment in August 2008, and is no longer
applicable. The amendments also correct errors in the titles of
analytical methods in DNPS and QCNPS TS 5.6.5, ``Core Operating Limits
Report (COLR),'' paragraph b. The proposed changes delete historical
analytical methods from DNPS and
[[Page 20641]]
QCNPS TS 5.6.5.b that are no longer applicable, and renumber the
remaining analytical methods.
Date of issuance: April 1, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 234/227, 246/241.
Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and
DPR-30. The amendments revised the Technical Specifications and
License.
Date of initial notice in Federal Register: June 30, 2009 (74 FR
31322). The October 5, 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 April 1, 2010.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: September 28, 2009, as supplemented by
letter dated January 20, 2010.
Brief description of amendment request: The proposed amendment
would support application of optimized weld overlays or full structural
weld overlays. Applying these weld overlays on the reactor coolant pump
suction and discharge nozzle dissimilar metal welds requires an update
to the DBNPS leak-before-break (LBB) evaluation.
Date of issuance: March 24, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 281.
Facility Operating License No. NPF-3: The amendment revised the
current licensing basis.
Date of initial notice in Federal Register: February 22, 2010 (75
FR 7628).
The January 20, 2010 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 amendment is contained
in a Safety Evaluation dated March 24, 2010.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: November 6, 2008; superseded by
letters dated August 4 and December 4, 2009.
Brief description of amendment: The amendment modifies the Crystal
River Unit 3 (CR-3) technical specifications (TS) surveillance
requirements (SRs) related to allowable voltage and frequency limits
for the emergency diesel generator (EDG) testing. Specifically, the
amendment revises the CR-3 TS SRs 3.8.1.2, 3.8.1.6, 3.8.1.10.c.3 and
3.8.1.10.c.4 to restrict the voltage and frequency limits for both slow
and fast EDG starts.
Date of issuance: December 10, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 236.
Facility Operating License No. DPR-72: Amendment revises the
facility operating license and the technical specifications.
Date of initial notice in Federal Register: September 8, 2009 (74
FR 46242). The supplement dated December 4, 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.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated December 10, 2009.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of application for amendment: March 4, 2009.
Brief description of amendment: The amendment changed the Duane
Arnold Energy Center Technical Specification (TS) Section 5.5.12
(Primary Containment Leakage Rate Testing Program) to exclude the Main
Steam pathway leakage contribution from the overall integrated leakage
rate Type A test measurement and from the sum of the leakage rates from
Type B and Type C tests and changed TS Section 3.6.1.3 (Primary
Containment Isolation Valves) to remove the repair criterion for main
steam isolation valves that fail their as-found leakage rate acceptance
criterion found in current Surveillance Requirement 3.6.1.3.9.
Date of issuance: March 31, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 276.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 30, 2009 (74 FR
31324).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 2010.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station (NMPNS), Unit No. 2 (NMP2), Oswego County, New
York
Date of application for amendment: June 29, 2009, as supplemented
on August 13, 2009, and February 3, 2010.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.5.12, ``10 CFR 50 Appendix J Testing Program
Plan,'' by replacing the reference to Regulatory Guide 1.163 with a
reference to Nuclear Energy Institute (NEI) topical report NEI 94-01,
Revision 2-A, as the implementation document used by NMPNS to develop
the NMP2 performance-based leakage testing program in accordance with
Option B of 10 CFR 50, Appendix J. In addition, the amendment allows
NMPNS to extend the current interval for the NMP2 primary containment
integrated leak rate test (ILRT) from 10 years to 15 years, and allows
successive ILRTs to be performed at 15-year intervals.
Date of issuance: March 30, 2010.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 134.
Renewed Facility Operating License No. NPF-069: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: October 20, 2009 (74 FR
53779).
The supplemental letters dated August 13, 2009, and February 3,
2010, 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 30, 2010.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station,
Unit No. 1, Salem County, New Jersey
Date of application for amendment: October 8, 2009, as supplemented
by letter dated February 25, 2010.
Brief description of amendments: The amendment approves a one-time
change to Technical Specification (TS) 6.8.4.i,
[[Page 20642]]
``Steam Generator (SG) Program,'' regarding the SG tube inspection and
repair required for the portion of the SG tubes passing through the
tubesheet region. Specifically, for Salem Unit No. 1 refueling outage
20 (planned for spring 2010) and subsequent operating cycles until the
next scheduled SG tube inspection, the amendment limits the required
inspection (and repair if degradation is found) to the portions of the
SG tubes passing through the upper 13.1 inches of the approximate 21-
inch tubesheet region. In addition, the amendment revises TS 6.9.1.10,
``Steam Generator Tube Inspection Report,'' to provide reporting
requirements specific to the one-time change.
Date of issuance: March 29, 2010.
Effective date: As of the date of issuance, to be implemented prior
to completion of refueling outage 20 (currently scheduled for spring
2010).
Amendment No.: 294.
Facility Operating License Nos. DPR-70 and DPR-75: The amendment
revised the TSs and the License.
Date of initial notice in Federal Register: January 5, 2010 (75 FR
464).
The letter dated February 25, 2010, provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the application beyond the scope
of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 29, 2010.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment requests: February 3, 2009, and March 3, 2009;
both applications were supplemented by letters dated November 20, 2009,
and January 20, 2010.
Brief description of amendments: The amendments approved a revision
to the South Texas Project (STP), Units 1 and 2 Fire Protection Program
for Fire Areas 27 and 31. In the event of a fire in the Fire Areas 27
and 31, the amendments allow the licensee to perform operator manual
actions to achieve and maintain safe shutdown in lieu of meeting the
circuit separation and protection requirements of Title 10 of the Code
of Federal Regulations, Part 50, Appendix R, Section III.G.2. The
amendments revised the License Condition 2.E, ``Fire Protection,'' in
the facility operating licenses, to reflect the changes. The approved
changes to the Fire Protection Program will be documented in the
licensee's ``Fire Hazards Analysis Report.''
Date of issuance: March 31, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 1--193; Unit 2--181.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses.
Date of initial notices in Federal Register: August 25, 2009 (74 FR
42929, 42930). The supplemental letters dated November 20, 2009, and
January 20, 2010, provided additional information that clarified the
applications, did not expand the scope of the applications as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2010.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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 51.12(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
application for amendment, (2) the amendment to
[[Page 20643]]
Facility Operating License, 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's (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].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland, and electronically on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected].
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 petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the 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.\1\ 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 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.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
requestor/petitioner seeks to adopt the contention of another
sponsoring requestor/petitioner, the requestor/petitioner who seeks to
adopt the contention must either agree that the sponsoring requestor/
petitioner shall act as the representative with respect to that
contention, or jointly designate with the sponsoring requestor/
petitioner a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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
[[Page 20644]]
representative, already holds an NRC-issued digital ID certificate).
Based upon this information, the Secretary will establish an electronic
docket for the hearing in this proceeding if the Secretary has not
already established an electronic docket.
Information about applying for a digital ID certificate is
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through EIE, users will be required to install a Web
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
social security numbers, home 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.
Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: March 22, 2010, as supplemented on March
23, 2010.
Description of amendment request: The previous Technical
Specification (TS) 3.4.17, ``Chemical and Volume Control System
(CVCS),'' Action B, allowed the licensee 24 hours to restore an
inoperable makeup water pathway from the Refueling Water Storage Tank
before taking further actions. This amendment increased the completion
time of TS 3.4.17, Action B, from 24 hours to 72 hours for fuel cycle
26.
Date of issuance: March 25, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 223.
Facility Operating License No. (DPR-23): Amendment revises the
technical specifications.
Public comments requested as to propose no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated
March 25, 2010.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Douglas A. Broaddus.
Dated at Rockville, Maryland, this 12th day of April 2010.
For The Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-8744 Filed 4-19-10; 8:45 am]
BILLING CODE 7590-01-P