[Federal Register Volume 73, Number 175 (Tuesday, September 9, 2008)]
[Notices]
[Pages 52412-52426]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-20567]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
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 staff) 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 August 14, 2008, to August 27, 2008. The
last
[[Page 52413]]
biweekly notice was published on August 26, 2008 (73 FR 50356).
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 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, 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 delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice,
person(s) may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
via electronic submission through the NRC E-Filing system for a hearing
and a petition for leave to intervene. 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. 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 within 60
days, 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 set forth the specific contentions which the petitioner/requestor
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
petitioner/requestor 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/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor 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 petitioner/requestor to relief. A petitioner/
requestor 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, and the Commission has not made a final
determination on the issue of no significant hazards consideration, 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.
[[Page 52414]]
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve 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 a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (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/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM is free and is
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. 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.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE 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
EIE 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 may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, 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.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
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, an Atomic Safety and Licensing Board, or
a 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. 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.
For further details with respect to this amendment action, 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 01F21, 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. 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 Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of amendment request: August 14, 2008.
Description of amendment request: The proposed amendment would
modify Specification 4.4.f.1, ``Containment Isolation Device
Verification,'' of the Technical Specifications (TS) to require
verification that the 36-inch containment purge and vent isolation
valves are sealed closed when the reactor is at greater than Cold
Shutdown conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
The Design Bases Accidents (DBA) that result in a release of
radioactive material within containment are a steam line break,
rupture of a rod cluster control assembly, and loss-of-coolant
accident (LOCA). In the analyses for each of these accidents, it is
assumed that containment isolation valves are either closed or
function to close within the required isolation time following
accident initiation. This ensures that potential leakage paths to
the environment
[[Page 52415]]
through containment isolation valves (including containment purge
and vent isolation valves) are minimized. The safety analyses assume
that the containment purge and vent isolation valves are closed at
accident initiation.
The safety function of the containment purge and vent isolation
valves is to support the Containment Isolation system by confining
fission products within the Primary Containment system boundary
during a DBA. The proposed amendment would require verification that
the containment purge and vent isolation valves are sealed closed
when the reactor is at greater than Cold Shutdown conditions. This
requirement ensures the valves are in their required DBA post-
accident position when the reactor is at greater than Cold Shutdown
conditions.
Verifying the containment purge and vent isolation valves are
sealed closed at 31-day intervals does not add, delete, or modify
any KPS system, structure, or component (SSC). Verifying that the
containment purge and vent isolation valves are sealed closed when
the reactor is at greater than Cold Shutdown conditions has no
adverse effect on the ability of the plant to mitigate the effects
of DBAs. The subject surveillance requirement constitutes a
verification of isolation valve position and has no effect on
equipment. Verification of valve closure only ensures the previous
assumptions made in evaluating the consequences of DBAs remain
valid. Therefore, there is no increase in the probability of an
accident by performing the surveillance in additional modes of plant
operation.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Verifying the containment purge and vent isolation valves are
sealed closed when the reactor is at greater than Cold Shutdown
conditions at 31-day intervals ensures these valves are in their
required DBA post-accident position when the design function is
required. The proposed amendment does not change the manner in which
these valves are operated when the reactor is at or below Cold
Shutdown or their design function. The proposed amendment does not
create any new failure mechanisms or malfunctions for plant
equipment or the nuclear fuel.
In addition, the containment purge and vent isolation valves are
not accident initiators. Their function is only for mitigation of
accidents.
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.
Verifying the containment purge and vent isolation valves are
sealed closed when the reactor is at greater than Cold Shutdown
conditions at 31-day intervals ensures these valves are in their
required DBA post-accident position when the design function is
required. The proposed amendment does not change the manner in which
these valves are operated when the reactor is at or below Cold
Shutdown condition.
The proposed amendment would align the KPS TS with applicable
NRC requirements stated in NUREG-0800 [``Standard Review Plan,''],
Section 6.2.4 and NUREG-0737 [``Clarification of Three Mile Island
Action Plan Requirements,''], Item II.E.4.2. The proposed amendment
does not result in altering or exceeding a design basis or safety
limit for the plant. The safety analysis of record, including
evaluations of the radiological consequences of design basis
accidents, will remain applicable and unchanged.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc.,
120 Tredegar Street, Richmond, VA 23219.
NRC Branch Chief: Lois James.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: August 1, 2008.
Description of amendment request: The proposed amendments would
authorize changes to the Updated Final Safety Analysis Report (UFSAR)
to account for small areas of carbon steel (CS) and low alloy steel
that may be exposed to the reactor coolant system (RCS).
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 Pressurizer vent nozzle and thermowell, as components of the
RCS, must maintain system pressure boundary. RCS design pressure is
2500 psig and design temperature is 670 [deg]F. The vent nozzle and
thermowell replacements are designed for the RCS pressure and
temperature. As described above, the material of the new Pressurizer
vent nozzle and thermowell is an improvement in the PWSCC [primary
water stress corrosion cracking] resistance of those components as
compared to the original components. The design of the new
Pressurizer vent nozzle and thermowell exposes small areas of the
Pressurizer shell carbon steel to a stagnant reactor coolant
environment. However, the corrosion of the Pressurizer shell is
considered negligible. Therefore, the replacement of the Pressurizer
vent nozzle and thermowell do not more than minimally increase the
likelihood of occurrence of a malfunction. Corrosion evaluations
performed show that all applicable ASME Code requirements are met.
It is concluded that the consequences of a Pressurizer vent
nozzle or Pressurizer thermowell failure resulting in a LOCA [loss-
of-coolant accident] are bounded by existing analysis. Therefore,
there is no increase in the probability or consequences of an
accident.
(2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No.
The only credible accident involving the failure of these
components is bounded by existing LOCA analyses. There are no new
accidents that need to be postulated due to the replacement of the
Pressurizer vent nozzle and Pressurizer thermowell. Therefore, this
proposed activity will not create the possibility of a new or
different kind of accident from any kind of accident previously
evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
No.
The mitigation technique selected for the Pressurizer vent
nozzle and the Pressurizer thermowell exposes a small area of CS to
the RCS environment. As required by the ASME Code, Section III, a
supporting corrosion evaluation was developed within each of the two
component designs. The technical package for the replacement of the
Pressurizer vent nozzle and the Pressurizer thermowell utilized
calculations to support the evaluation of the acceptability of this
repair/replacement activity. The corrosion evaluation for the
Pressurizer vent resulted in a conservative general stagnant
corrosion rate of 0.0018 inches per year and the corrosion
evaluation for the Pressurizer thermowell resulted in a conservative
general corrosion rate of 0.00142 inches per year. The critical
corrosion distance is the radius from the exposed CS surface to the
edge of the weld pad. This distance is at least 1.1 inches for both
the vent and thermowell designs. With this distance, a corrosion
rate of less than 2 mils per year is not significant when compared
to the 60 year component design life, which begins at the time of
installation.
The original Pressurizer was designed to meet Section III of the
ASME Code, and the Pressurizer, as modified, meets Section III of
the ASME code. Although this change does expose small areas of CS in
the Pressurizer, the 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
[[Page 52416]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie C. Wong.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of amendment request: July 8, 2008.
Description of amendment request: The proposed amendment to Indian
Point Nuclear Generating Units Nos. 2 and 3 (IP2 and IP3) would require
the licensee to submit information and analyses associated with
extending the Reactor Vessel (RV) Inservice Inspection (ISI) Interval
from 10 to 20 years for specific pressure retaining welds in the RV.
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 will revise the license to
require the submission of information and analyses to the NRC
following completion of each ASME [American Society of Mechanical
Engineers] [C]ode, Section XI, Category B-A and B-D Reactor Vessel
weld inspection. The extension of the ISI from 10 to 20 years is
being evaluated as part of the relief request independent from the
license change. Submission of the information and analyses can have
no effect on the consequences of an accident or the probability of
an accident because the submission of information is not related to
the operation of the plant or any equipment, the programs and
procedures used to operate the plant, or the evaluation of
accidents. The submittal of information and analyses provides the
opportunity for the NRC to independently assess the information and
analyses.
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 will only affect the
requirement to submit information and analyses when specified
inspections are performed. There are no changes to plant equipment,
operating characteristics or conditions, programs, and procedures or
training. Therefore, there are no potential new system interactions
or failures that could 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 change will revise the license to
require the submission of information and analyses to the NRC
following completion of each ASME [C]ode, Section XI, Category B-A
and B-D Reactor Vessel weld inspection which does not affect any
Limiting Conditions for Operation used to establish the margin of
safety. The requirement to submit information and analyses is an
administrative tool to assure the NRC has the ability to
independently review information developed by the [l]icensee. The
proposed change does not involve a significant reduction in [a]
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: July 9, 2008.
Description of amendment request: The proposed amendment will
revise the test acceptance criteria specified in the Technical
Specification Surveillance Requirement (SR) 3.8.1.10 for the Diesel
Generator (DG) endurance test. The load ranges and power factors
specified for the test will be changed for consistency with the
associated 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. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change revises the acceptance criteria to be
applied to an existing surveillance test of the facility emergency
diesel generators (DGs). Performing a surveillance test is not an
accident initiator and does not increase the probability of an
accident occurring. The proposed new acceptance criteria will assure
that the DGs are capable of carrying the peak electrical loading
assumed in the various existing safety analyses which take credit
for the operation of the DGs. Establishing acceptance criteria that
bound existing analyses validates the related assumption used in
those analyses regarding the capability of equipment to mitigate
accident conditions. Therefore the proposed change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change revises the test acceptance criteria for
a specific performance test conducted on the existing DGs. The
proposed change does not involve installation of new equipment or
modification of existing equipment, so no new equipment failure
modes are introduced. The proposed revision to the DG surveillance
test acceptance criteria also is not a change to the way that the
equipment or facility is operated and no new accident initiators are
created. Therefore the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The conduct of performance tests on safety-related plant
equipment is a means of assuring that the equipment is capable of
maintaining the margin of safety established in the safety analyses
for the facility. The proposed change in the DG technical
specification surveillance test acceptance criteria is consistent
with values assumed in existing safety analyses is consistent with
the design rating of the DGs. Therefore the propose change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: May 5, 2008.
Description of amendment request: The proposed amendment would
correct an error in Section A.1 of the renewed operating license and
remove several outdated license conditions relating to surveillance
requirements. Specifically, it would remove the words ``filed by
Entergy Nuclear Palisades, LLC (ENP) and Entergy Nuclear
[[Page 52417]]
Operations, Inc. (ENO)'' in Section A.1, spell-out acronyms used in
Section 1.F, and delete license conditions 2.C.(4) and 2.C.(5), and
delete Table 2.C.(5).
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 license amendment deletes incorrect or outdated
information from the renewed facility operating license. The
proposed amendment does not involve operation of the required
structures, systems or components (SSCs) in a manner or
configuration different from those previously recognized or
evaluated.
Modification of renewed facility operating license sections 1.A
and 1.F and deletion of license conditions 2.C.(4), 2.C.(5), and
Table 2.C.(5) is administrative and has no impact on plant operation
or equipment.
Therefore, operation of the facility in accordance with the
proposed amendment would 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 a physical
alteration of any SSC or change the way any SSC is operated. The
proposed license amendment does not involve operation of any
required SSCs in a manner or configuration different from those
previously recognized or evaluated.
Modification of renewed facility operating license sections 1.A
and 1.17 and deletion of license conditions 2.C.(4), 2.C.(5), and
Table 2.C.(5) is administrative and has no impact on plant operation
or equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Modification of renewed facility operating license sections 1.A
and 1.F and deletion of license conditions 2.C.(4), 2.C.(5), and
Table 2.C.(5) is administrative and has no impact on plant operation
or equipment or on any margin of safety.
Therefore, the proposed amendment would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: Lois M. James.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: May 5, 2008
Description of amendment request: The proposed amendment would
revise renewed facility operating license DPR-20 to remove license
condition 2F. This license condition describes reporting requirements
for exceeding the facility steady-state reactor core power level
described in condition 2.C.(1). The proposed change is consistent with
the Nuclear Regulatory Commission (NRC)-approved change notice
published in the Federal Register on November 4, 2005, announcing the
availability of this improvement through the consolidated line item
improvement process. The Federal Register Notice included a model
safety evaluation and model no significant hazards consideration (NSHC)
determination, relating to the elimination of the license condition
involving reporting of violations of other requirements (typically in
License Conditions 2.C) in the operating license of some commercial
nuclear power plants. The licensee affirmed the applicability of the
model NSHC determination in its application dated May 5, 2008.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and therefore does not significantly increase the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the 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 change deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: Lois M. James.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood; Station, Units 1 and 2, Will County, Illinois; Docket Nos.
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois
Date of amendment request: July 29, 2008.
Description of amendment request: The proposed amendments would
remove time, cycle, or modification-related items from the operating
licenses (OLs) and technical specifications (TSs) at both stations.
Additionally, the proposed amendments would correct typographical
errors introduced into the TSs at both stations in previous amendments.
The time, cycle, or modification-related items have been implemented or
superseded, are no longer applicable, and no longer need to be
maintained in their associated OLs or TSs.
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 initial conditions and methodologies used in the accident
analyses remain unchanged. The proposed changes do not change or
alter the design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, accident
analyses results are not impacted.
All changes proposed by EGC in this amendment request are
administrative in nature, and are removing one-time requirements
that have been satisfied or items that are no longer applicable.
There are no physical changes to the facilities, nor any changes to
the station operating procedures,
[[Page 52418]]
limiting conditions for operation, or limiting safety system
settings.
Based on the above discussion, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
None of the proposed changes affect the design or operation of
any system, structure, or component in the plant. The safety
functions of the related structures, systems, or components are not
changed in any manner, nor is the reliability of any structure,
system, or component reduced by the revised surveillance or testing
requirements. The changes do not affect the manner by which the
facility is operated and do not change any facility design feature,
structure, system, or component. No new or different type of
equipment will be installed. Since there is no change to the
facility or operating procedures, and the safety functions and
reliability of structures, systems, or components are not affected,
the proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Based on this evaluation, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to the Facility Operating Licenses and TS
are administrative in nature and have no impact on the margin of
safety of any of the TS. There is no impact on safety limits or
limiting safety system settings. The changes do not affect any plant
safety parameters or setpoints. The Operating License Conditions
have been satisfied as required. There are no changes to the
conditions themselves.
Based on this evaluation, the proposed change does not involve a
significant reduction in a margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Florida Power and Light Company, et al. , Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: January 23, 2008.
Description of amendment request: Replace the current Technical
Specification pressure/temperature (P/T) limit curves with new P/T
limit curves applicable to 55 effective full-power years (EFPY). The
low-temperature overpressure protection (LTOP) requirements, which are
based on the P/T limits, will also be applicable to 55 EFPY.
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) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes have been determined in accordance with the
methodologies set forth in the regulations to provide an adequate
margin of safety to ensure that the reactor vessel will withstand
the effects of normal startup and shutdown cyclic loads due to
system temperature and pressure changes as well as the loads
associated with reactor trips. The regulations of 10 CFR Part 50
Appendix A, Design Criterion 14 and Design Criterion 31 remains
satisfied. The pressure-temperature (P/T) limit curves in the
Technical Specifications are conservatively generated in accordance
with the fracture toughness requirements of the ASME [American
Society of Mechanical Engineers] Code Section XI, Appendix G. The
margins of safety against fracture provided by the P/T limits using
the requirements of 10 CFR 50 Appendix G are equivalent to those
recommended in ASME Section XI, Appendix G. The Adjusted Reference
Temperature (ART) values are based on the guidance of RG [Regulatory
Guide] 1.99 [Reference 4].
The proposed changes will not result in physical changes to
structures, systems or components SSCs or to event initiators or
precursors. Changing the heatup and cooldown curves and the pressure
relief setpoints to reflect 55 EFPY does not affect the ability to
control the RCS [reactor coolant system] at low temperatures such
that the integrity of the reactor coolant pressure boundary would
not be compromised by violating the P/T limits.
The proposed changes will not impact assumptions and conditions
previously used in the radiological consequence evaluations nor
affect mitigation of these consequences due to an accident described
in the UFSAR [Updated Final Safety Analysis Report]. Also, the
proposed changes will not impact a plant system such that previously
analyzed SSCs might be more likely to fail. The initiating
conditions and assumptions for accidents described in the UFSAR
remain as analyzed.
Thus, based on the above, reasonable assurance is provided that
the proposed amendment does not significantly increase the
probability or consequences of accidents previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The requirements for P/T limit curves and LTOP have been in
place since the beginning of plant operation. The revised curves are
based on a later edition of Section XI of the ASME Code that
incorporates current industry standards for P/T curves. The revised
curves also are based on reactor vessel irradiation damage
predictions using RG 1.99 methodology. No new failure modes are
identified nor are any SSCs required to be operated outside of their
design bases. Consequently, the proposed amendment does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed P/T curves continue to maintain the safety margins
of 10 CFR 50 Appendix G by defining the limits of operation which
prevent nonductile failure of the reactor pressure vessel. Analyses
have demonstrated that the fracture toughness requirements are
satisfied and that conservative operating restrictions are
maintained for the purpose of low temperature overpressure
protection. The P/T limit curves provide assurance that the RCS
pressure boundary will behave in a ductile manner and that the
probability of a rapidly propagating fracture is minimized.
Therefore, operation in accordance with the proposed amendment would
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: April 4, 2008.
Description of amendment request: The licensee proposed to change
the Technical Specifications (TS) to revise requirements for
unavailable barriers by adding Limiting Condition for Operation (LCO)
3.0.9. This LCO would establish conditions under which systems would
remain operable when required physical barriers are not capable of
providing their related support function. This proposed amendment is
consistent with the NRC's approved Technical Specification Task
[[Page 52419]]
Force (TSTF) Improved Standard Technical Specifications Change
Traveler, TSTF-427, Revision 2. A notice of availability of this TS
improvement was published in the Federal Register on October 3, 2006
(71 FR 58444) as part of NRC's Consolidated Line Item Improvement
Process (CLIIP).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided an
analysis of the issue of no significant hazards consideration by citing
the proposed NSHC determination published by the NRC staff in the
Federal Register referenced above. That proposed NSHC is reproduced
below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an unavailable barrier if risk is assessed and managed.
The postulated initiating events which may require a functional
barrier are limited to those with low frequencies of occurrence, and
the overall TS system safety function would still be available for
the majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on the
allowance provided by proposed LCO 3.0.9 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.9.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by this change. The addition of a
requirement to assess and manage the risk introduced by this change
will further minimize possible concerns. Therefore, this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to an unavailable barrier, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an unavailable
barrier, if risk is assessed and managed. The postulated initiating
events which may require a functional barrier are limited to those
with low frequencies of occurrence, and the overall TS system safety
function would still be available for the majority of anticipated
challenges. The risk impact of the proposed TS changes was assessed
following the three-tiered approach recommended in RG [Regulatory
Guide] 1.177. A bounding risk assessment was performed to justify
the proposed TS changes. This application of LCO 3.0.9 is predicated
upon the licensee's performance of a risk assessment and the
management of plant risk. The net change to the margin of safety is
insignificant as indicated by the anticipated low levels of
associated risk (ICCDP [incremental conditional core damage
probability] and ICLERP [incremental conditional large early release
probability] ) as shown in Table 1 of Section 3.1.1 in the Safety
Evaluation. Therefore, this change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis cited by the licensee, and
has found that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the proposed
amendment involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Lois M. James.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: April 22, 2008.
Description of amendment request: The licensee proposed to change
the Technical Specifications (TS) to (1) revise the surveillance
requirement frequency in Specification 3.1.3, ``Control Rod
Operability,'' to require control rod notch testing to be performed at
a 31-day frequency for both partially and fully withdrawn control rods;
and (2) revise Example 1.4-3 in Section 1.4, ``Frequency,'' to clarify
the applicability of the 1.25 surveillance test interval extension.
These proposed changes are consistent with the NRC's approved Technical
Specification Task Force (TSTF) Improved Standard Technical
Specifications (STS) Change Traveler, TSTF-475, Revision 1. A notice of
availability of this TS improvement was published in the Federal
Register on November 13, 2007 (72 FR 63935), as part of the NRC's
Consolidated Line Item Improvement Process (CLIIP).
Basis for proposed no significant hazards consideration
determination: As required by 10 FR 50.91(a), the licensee provided an
analysis of the issue of no significant hazards consideration by citing
the proposed NSHC determination published by the NRC staff in the
Federal Register notice referenced above. That proposed NSHC is
reproduced below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM [Source Range
Monitoring] Insert Control Rod Action.'' TSTF-475, Revision 1,
modifies NUREG-1433 (BWR [Boiling Water Reactor]/4) and NUREG-1434
(BWR/6) STS. The changes (1) revise TS testing frequency for
surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod
OPERABILITY,'', and (2) revise Example 1.4-3 in Section 1.4
``Frequency'' to clarify the applicability of the 1.25 surveillance
test interval extension. The consequences of an accident after
adopting TSTF-475, Revision 1 are no different than the consequences
of an accident prior to adoption. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Accident Previously
Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change will not introduce new failure modes or effects and
will not, in the absence of other unrelated failures, lead to an
accident whose consequences exceed the consequences of accidents
previously analyzed. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2
frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' and (2) revise
Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the
applicability of the 1.25 surveillance test interval extension. The
GE Nuclear Energy Report, ``CRD Notching Surveillance Testing for
Limerick Generating Station,'' dated November 2006, concludes that
extending the control rod notch test interval from weekly to monthly
is not expected to impact the reliability of the scram system and
that the analysis supports the decision to change the surveillance
frequency. Therefore, the proposed changes in TSTF-475, Revision 1
[[Page 52420]]
are acceptable and do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the analysis cited by the licensee, and
has found that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the proposed
amendment involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Lois M. James.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: June 26, 2008.
Description of amendment request: The proposed amendments would
amend the Facility Operating Licenses by revising the licensing basis
loss of coolant accident and main steam line break accident
radiological dose consequences for Prairie Island Nuclear Generating
Plant, Units 1 and 2, as currently described in the Updated Safety
Analysis Report Section 14.5 and Section 14.9. This proposed amendment
also proposes concomitant amendments to Appendix A of the Facility
Operating Licenses, Technical Specifications (TS) 3.3.5, ``Containment
Ventilation Isolation Instrumentation'', 3.4.17, ``RCS [Reactor Coolant
System] Specific Activity'', and 3.6.3, ``Containment Isolation
Valves'', which are necessary to implement the proposed revised
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. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes implementing revised
loss of coolant accident and main steam line break accident dose
consequence analyses to address modeling nonconservatisms and update
the analyses for new fuel types and provide margin for power
uncertainty. These analyses assumed that the containment inservice
purge system penetrations are isolated, thus this license amendment
request proposes Technical Specification revisions which will
require these penetrations to be blind flanged during plant
operations; these changes allow the Technical Specification
requirements for containment ventilation isolation instrumentation
to be removed. This license amendment request also proposes
associated more restrictive limits in the Technical Specification
for reactor coolant system specific activity since the main steam
line break accident analysis assumed lower limits.
The accident radiological dose consequences analyses inputs,
methodologies and outputs modified by this request are not accident
initiators and do not affect the frequency of occurrence of
previously analyzed transients. Likewise, the reactor coolant system
specific activity limits are not accident initiators and do not
affect the frequency of occurrence of previously analyzed
transients.
The containment inservice purge system is not an accident
initiator and therefore removal of its Technical Specifications does
not involve an increase in the probability of an accident. The
Technical Specification changes proposed in this license amendment
request require the containment inservice purge system to be blind
flanged during Modes 1, 2, 3, and 4, therefore removal of the
containment ventilation isolation instrumentation Technical
Specifications and other Technical Specification system operating
requirements does not involve an increase in the consequences of an
accident previously evaluated.
The loss of coolant accident and main steam line break accident
radiological dose consequences analyses demonstrated the results are
within the applicable regulatory limits and guidance using revised
inputs, including the proposed lower Technical Specification reactor
coolant system specific activity limits, and methodologies. Thus
these changes do not involve a significant increase in the
consequences of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This license amendment request proposes implementing revised
loss of coolant accident and main steam line break accident dose
consequence analyses to address modeling nonconservatisms and update
the analyses for new fuel types and provide margin for power
uncertainty. These analyses assumed that the containment inservice
purge system penetrations are isolated, thus this license amendment
request proposes Technical Specification revisions which will
require these penetrations to be blind flanged during plant
operations; these changes allow the Technical Specification
requirements for containment ventilation isolation instrumentation
to be removed. This license amendment request also proposes
associated more restrictive limits in the Technical Specification
for reactor coolant system specific activity since the main steam
line break accident analysis assumed lower limits.
This license amendment request does not involve physical changes
to the plant structures, systems or components and there is no
adverse impact on component or system interactions due to the
proposed changes. The modes of operation of the plant remain
unchanged and the design functions of the safety systems remain in
compliance with the applicable safety analysis acceptance criteria.
These changes do not create new failure modes or mechanisms and no
new accident precursors are generated.
When the containment inservice purge system is not being
operated, current Technical Specifications require the system's
penetrations to be blind flanged in Modes 1, 2, 3, and 4 to provide
post-accident containment integrity. This license amendment proposes
to require the system penetrations to be blind flanged at all times
during these Modes and prevent operation of the system in these
Modes. Since containment integrity is provided with the penetrations
blind flanged and this change only extends the time during which the
system is in this configuration, these changes do not create the
possibility of a new or different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This license amendment request proposes implementing revised
loss of coolant accident and main steam line break accident dose
consequence analyses to address modeling nonconservatisms and update
the analyses for new fuel types and provide margin for power
uncertainty. These analyses assumed that the containment inservice
purge system penetrations are isolated, thus this license amendment
request proposes Technical Specification revisions which will
require these penetrations to be blind flanged during plant
operations; these changes allow the Technical Specification
requirements for containment ventilation isolation instrumentation
to be removed. This license amendment request also proposes
associated more restrictive limits in the Technical Specification
for reactor coolant system specific activity since the main steam
line break accident analysis assumed lower limits.
The loss of coolant accident and main steam line break accident
radiological dose consequences analyses have incorporated revised
inputs, including the proposed lower Technical Specification reactor
coolant system specific activity limits, and utilized revised
methodologies. The results of these revised analyses satisfy the
applicable regulatory limits and guidance. There is no adverse
effect on plant safety due to this proposed license amendment.
The containment inservice purge system is not credited for
mitigation of any accidents or any other safety function, thus,
removal of its associated Technical Specifications does not involve
reduction in a margin of safety. The containment ventilation
isolation instrumentation system is credited for isolation of the
containment inservice purge system following an accident and the
valves are assumed to meet containment integrity
[[Page 52421]]
leakage rate limits. This license amendment request proposes to
require the containment inservice purge system containment
penetrations to be blind flanged during Modes 1, 2, 3, and 4 and the
blind flanged penetrations will be required to meet containment
integrity leakage rate limits. With these changes, containment
integrity is maintained in accordance with the current Technical
Specification requirements, thus, this change does not involve
reduction in a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: July 30, 2008.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.8.3, ``Onsite Power Distribution
Systems,'' to establish a separate TS Action statement for inoperable
inverters associated with the 120 volt alternating current (VAC)
distribution panels. The intent of the proposed amendment is to extend
the allowed outage time for inoperable inverters from 8 hours to 24
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 inverters and associated 120 VAC distribution panels are not
initiators to any accident sequence analyzed in the Updated Final
Safety Analysis Report (UFSAR).
The proposed change does not increase the number of inverters
permitted to be inoperable at one time. With one or both inverters
inoperable in a single channel, sufficient capacity and capability
remain to assure required safety functions can be performed. The
proposed changes do not involve any physical change to structures,
systems, or components (SSCs) and do not alter the method of
operation or control of SSCs. The current assumptions in the safety
analysis regarding accident initiators and mitigation of accidents
are unaffected by these proposed changes. The likelihood of
previously analyzed failures remains unchanged.
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.
No physical changes will be made to the plant or how the plant
is operated. As such, no new or different kind of accident due to a
credible new failure mechanism, malfunction, or accident initiator
will be created as a result of this proposed change. Any alteration
in procedures will continue to ensure that the plant remains within
analyzed limits, and no change is required to the procedures relied
upon to respond to an off-normal event as described in the UFSAR.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change would extend the allowed outage time for one
or two inoperable inverters in a single channel. The proposed change
does not increase the number of inverters permitted to be inoperable
at one time. There is no change to any design basis or safety
limits. Operation in accordance with the proposed TS ensures that
the 120 VAC instrument distribution system is capable of performing
its functions as described in the UFSAR.
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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
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, September 12, November 19, December 13, and
December 17, 2007; January 10 (4 letters), January 11 (4 letters),
January 14, and January 18 (5 letters), January 31, February 25 (2
letters), March 5, March 10 (2 letters), March 25, March 27, April 4,
April 24, April 29, May 15, May 20, May 21, July 10, and July 16, 2008.
Brief description of amendment: The amendment increased the
Millstone Power Station, Unit No. 3 (MPS3) maximum steady-state reactor
core power level from the previous licensed thermal power level of
3,411 megawatts thermal (MWt) to 3,650 MWt, which is an increase of
approximately 7 percent. The amendment revises the MPS3 Operating
License and Technical Specifications necessary to implement the
increased power level.
Date of issuance: August 12, 2008.
Amendment No.: 242.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Facility Operating License No. NPF-49: Amendment revised the
License and Technical Specifications.
Date of individual notice of issuance in Federal Register: August
20, 2008 (73 FR 49222).
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: June 17, 2008.
Brief description of amendment request: The proposed amendment
would revise Technical Specification (TS) 5.5.9, Steam Generator (SG)
Program, and TS 5.6.9, Steam Generator Tube Inspection Report. For TS
5.5.9, the amendment would incorporate a one-cycle interim alternate
repair criteria in the provisions for SG tube
[[Page 52422]]
repair criteria during Byron, Unit No. 2, refueling outage 14 and the
subsequent operating cycle. For TS 5.6.9, the amendment would revise
the current reporting requirements. The proposed changes only affect
Byron, Unit No. 2; however, they are docketed for both Byron units
because the TSs are common to both units.
Date of publication of individual notice in Federal Register:
August 5, 2008 (73 FR 45485).
Expiration date of individual notice: September 5, 2008 (public
comment), October 5, 2008 (hearing requests).
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 Systems (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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of application for amendment: May 10, 2007, as supplemented by
letters dated January 10 and July 18, 2008.
Brief description of amendment: The amendment describes the long-
term coupon surveillance program for the carborundum samples found in
the Unit No. 1 spent fuel pool (SFP). The program verifies that the
carborundum degradation rates assumed in the licensee's analyses to
prove subcriticality, as required by Title 10 of the Code of Federal
Regulations, Section 50.68, remain valid over the 70-year life span of
the Unit No. 1 SFP.
Date of issuance: August 27, 2008.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 288.
Renewed Facility Operating License No. DPR-53: Amendment revised
the License and fulfills the requirements identified in Appendix C,
Additional Conditions, to Renewed Facility Operating License No. DPR-53
as further described in Amendment No. 267 issued on June 3, 2004.
Date of initial notice in Federal Register: June 19, 2007 (72 FR
33780).
The letters dated January 10 and July 18, 2008, 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 August 27, 2008.
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, Will County, Illinois; Exelon
Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron
Station, Unit Nos. 1 and 2, Ogle County, Illinois; Exelon Generation
Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1,
DeWitt County, Illinois; Exelon Generation Company, LLC, Docket Nos.
50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy
County, Illinois; Exelon Generation Company, LLC, Docket Nos. 50-373
and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County,
Illinois; AmerGen Energy Company, LLC, et al., Docket No. 50-219,
Oyster Creek Nuclear Generating Station, Ocean County, New Jersey;
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania; Exelon Generation Company, LLC,
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois; AmerGen Energy Company, LLC,
Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin
County, Pennsylvania
Date of application for amendment: July 19, 2007, as supplemented
on July 7, 2008.
Brief description of amendment: The amendments will update the
requirements in the Technical Specifications (TS) 5.3.1 ``Facility
Staff Qualifications,'' or TS 6.3.1, ``Unit Staff Qualifications,''
that have been outdated based on licensed operator training programs
accredited by the National Academy for Nuclear Training Academy
Document, ACAD 00-003, Revision 1, dated April 2004, and the revised
Title 10 of the Code of Federal Regulations, Part 55, ``Operators'
Licenses.''
Date of issuance: July 25, 2008.
Effective Date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 152, 152, 156, 156, 180, 228, 220, 189, 176, 267,
267, 271, 240, 235, 265
Facility Operating License Nos. NPF-72, NPF-77, NPF-37 and NPF-66,
NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-16, DPR-55, DPR-56, DPR-29,
DPR-30 and DPR-50: The amendments revised the Technical Specifications
and License.
Date of initial notice in Federal Register: December 4, 2007 (72 FR
68214). The supplemental letter contained clarifying information, did
[[Page 52423]]
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 25, 2008.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-282, Prairie Island
Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota
Date of application for amendment: August 16, 2007, as supplemented
by letter dated June 13, 2008.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) for Prairie Island Nuclear Generating Plant, Unit
1. The amendment revises TS 3.8.1 ``AC Sources--Operating'' to require
monthly testing of the Unit 1 emergency diesel generators at or above
2500 kilowatts.
Date of issuance: August 15, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 187.
Facility Operating License No. DPR-42: Amendment revises the TSs.
Date of initial notice in Federal Register: January 28, 2008 (73 FR
5226).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in Safety Evaluation dated August 15, 2008.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: August 16, 2007, as supplemented
by letter dated June 16, 2008.
Brief description of amendment: The amendment revises the Technical
Specification (TS) requirements related to control room envelope
habitability in TS 3.7.9, ``Control Room Emergency Air Treatment System
(CREATS),'' and TS Section 5.5, ``Programs and Manuals.'' The changes
are consistent with the Nuclear Regulatory Commission approved
Industry/Technical Specification Task Force Traveler No. 448, Revision
3. The availability of this TS improvement was published in the Federal
Register on January 17, 2007 (72 FR 2022), as part of the consolidated
line item improvement process.
Date of issuance: August 27, 2008.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 105.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: October 23, 2007 (72 FR
60035).
The June 16, 2008, supplemental letter 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 August 27, 2008.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket No. 50-362, San
Onofre Nuclear Generating Station, Unit 3, San Diego County, California
Date of application for amendments: September 24, 2007, as
supplemented by letters dated February 22 and March 27, 2008.
Brief description of amendments: Approves the revision to the SONGS
3 Technical Specification 5.5.2.15, ``Containment Leakage Rate Testing
Program,'' of a one-time extension from the currently approved 15-year
interval since the last Integrated Leak Rate Test to a 16-year
interval.
Date of issuance: August 15, 2008.
Effective date: to be implemented within 60 days of issuance.
Amendment No.: Unit 3-210.
Facility Operating License No. NPF-15: The amendments revised the
Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: October 23, 2007 (72 FR
60036). The supplements dated February 22 and March 27, 2008, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the U.S. Nuclear Regulatory Commission staff 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 August 15, 2008.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: March 26, 2008.
Brief description of amendment: The proposed amendment would revise
the Updated Final Safety Analysis Report (UFSAR) to reflect approval to
use the Boiling Water Reactor Vessel and Internals Project reactor
pressure vessel integrated surveillance program as the basis for
demonstrating the compliance with the requirements of Appendix H to
Title 10 of the Code of Federal Regulations Part 50, ``Reactor Vessel
Material Surveillance Program Requirements.''
Date of issuance: August 14, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 273.
Renewed Facility Operating License No. DPR-33: Amendment revised
the UFSAR.
Date of initial notice in Federal Register: June 3, 2008 (73 FR
31723).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 14, 2008.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: August 20, 2007, as supplemented
by letter dated March 12, 2008.
Brief description of amendment: The amendment revised Technical
Specification 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,''
and its associated Surveillance Requirement 3.8.3.1 to increase the
current minimum emergency diesel generator (EDG) fuel oil inventory
required to be maintained onsite. The increase in minimum EDG fuel oil
would provide conservative margin against potential vortex effects that
could occur during fuel oil transfer pump operation.
Date of issuance: August 27, 2008.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 185.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51866). The supplemental letter dated March 12, 2008, provided
additional
[[Page 52424]]
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 as published in
the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 27, 2008.
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 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, person(s) may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request via electronic submission
through the NRC E-Filing system for a hearing and a petition for leave
to intervene. 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
[[Page 52425]]
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 petitioner/requestor 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
petitioner/requestor 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 petitioner/requestor
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
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor 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.
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve 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 a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (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/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM is free and is
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. 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.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE 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
EIE 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 may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, 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
[[Page 52426]]
depositing the document with the provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
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, an Atomic Safety and Licensing Board, or
a 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. 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.
Exelon Generation Company, LLC, Docket No. 50-249, Dresden Nuclear
Power Station, Unit 3, Grundy County, Illinois
Date of amendment request: August 18, 2008.
Description of amendment request: The amendment revises Technical
Specification 3.4.5, ``RCS Leakage Detection Instrumentation,'' to
support implementation of an alternative method of verifying that
unidentified leakage in the drywell is within limits.
Date of issuance: August 22, 2008.
Effective date: As of the date of issuance and shall be implemented
by 12:00 pm CDT on August 24, 2008.
Amendment No.: 221.
Facility Operating License No. DPR-25: Amendment revises the
technical specifications and the operating license.
Public comments requested as to proposed no significant hazards
consideration (NSHC):
No. On August 17, 2008, the staff issued a Notice of Enforcement
Discretion, which was effective immediately and remained in effect
until this amendment was issued.
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 August 22,
2008.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation.
NRC Branch Chief: Russell Gibbs.
Dated at Rockville, Maryland, this 29th day of August 2008.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E8-20567 Filed 9-8-08; 8:45 am]
BILLING CODE 7590-01-P