[Federal Register Volume 72, Number 232 (Tuesday, December 4, 2007)]
[Notices]
[Pages 68206-68224]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-23225]
-----------------------------------------------------------------------
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 November 8, 2007 to November 21, 2007. The
last biweekly notice was published on November 20, 2007 (72 FR 65360).
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. Within 60 days after the date of publication of this
notice, the licensee 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 for a hearing and a petition for leave to intervene.
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
[[Page 68207]]
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.
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 on 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 Viewer(tm) to
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ 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
[[Page 68208]]
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].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County,
North Carolina
Date of amendments request: January 22, 2007, as supplemented by
letters dated June 21, July 18, July 31, and October 15, 2007.
Description of amendments request: The amendment would revise the
Technical Specifications to support the transition to AREVA NP fuel and
core design methodologies.
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 amendments revise the list of NRC-approved
analytical methods used to establish core operating limits. Core
operating limits are established to ensure that fuel design limits
are not exceeded during operating transients or accidents. The
analytical methods used to determine core operating limits are those
methods that have previously been found acceptable by the NRC and
are required to be listed in the Technical Specification section
governing the Core Operating Limits Report. The application of these
NRC-approved analytical methods will continue to ensure that
acceptable operating limits are established and applied to operation
of the reactor core.
The proposed amendments will add a new Technical Specification
3.2.3, ``Linear Heat Generation Rate (LHGR),'' for fuel bundles, add
a new definition to Technical Specification 1.1 for LHGR, and revise
Technical Specifications 3.4.1 and 3.7.6 to incorporate restrictions
on LHGR when in single recirculation loop operation or with an
inoperable Turbine Bypass System. These LHGR limits will be
established using NRC-approved analytical methods to ensure that
fuel performance during normal, transient, and accident conditions
is acceptable.
Based on the above, the proposed amendments do not involve an
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
As previously stated, the proposed amendments support transition
from Global Nuclear Fuels Americas (GNF-A) fuel and core design and
analysis services to AREVA NP fuel and core design and analysis
services. The AREVA NP fuel assemblies which will be used in the
BSEP Unit 1 and 2 cores will be similar in design to the GNF-A fuel
that will be co-resident in the cores. The BSEP, Unit 1 and 2 cores
in which this fuel will operate will be designed to meet all
applicable design and licensing criteria. Adherence to these design
and licensing criteria will not introduce any new modes of operation
or introduce any new accident precursors, and thus will preclude the
introduction 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 amendments will continue to require that core
operating limits be determined using NRC-approved analytical
methods. Acceptable fuel performance is obtained by ensuring that
the peak cladding temperature (PCT) during a postulated design basis
loss-of-coolant accident (LOCA) is maintained less than the limits
specified in 10 CFR 50.46, and that the core remains in a coolable
geometry following a postulated design basis LOCA. The proposed
amendments ensure that adequate margin will continue to be
maintained to the 2200 degree PCT limit of 10 CFR 50.46, and the use
of NRC-approved analytical methods will continue to ensure
acceptable fuel performance during normal operations, as well as
during transient and accident conditions. Therefore, the proposed
amendments do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: August 6, 2007.
Description of amendments request: The amendment would revise the
Technical Specifications (TSs) to implement Technical Specification
Task Force (TSTF) Change TSTF-343, Revision 1, which allows the
performance of visual examinations of the primary containment to be
performed in accordance with the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code, Section XI,
Subsections IWE and IWL. The amendment would also make an
administrative change to the TSs by eliminating a one-time requirement
to perform containment leak rate testing that has already been
completed.
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 affects the frequency of visual examinations
that will be performed for the concrete surfaces of the containment
for the purpose of the Primary Containment Leakage Rate Testing
Program. In addition, the proposed change allows those examinations
to be performed during power operation as opposed to during a
refueling outage. The frequency of visual examinations of the
metallic and concrete surfaces of the containment and the mode of
operation during which those examinations are performed has no
relationship to or adverse impact on the probability of any of the
initiating events assumed in the accident analyses. The proposed
change would allow
[[Page 68209]]
visual examinations that are performed in accordance with NRC-
approved ASME Section XI Code requirements, except where relief has
been granted by the NRC, to meet the intent of visual examinations
specified by Regulatory Guide 1.163, without requiring additional
visual examinations in accordance with the Regulatory Guide. The
intent of early detection of deterioration will continue to be met
by the more vigorous requirements of the Code-required visual
examinations. As such, the safety function of the containment as a
fission product barrier is maintained.
The proposed change also includes the removal of an item in TS
5.5.12 which was incorporated to establish deadlines for performing
the performance-based Type A leakage tests in conjunction with
changing, on a one-time basis, the Type A test frequency. The
specified Unit 1 and Unit 2 Type A test have been completed. As
such, removal of this item is an administrative change.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. The proposed change does not involve the addition or removal
of any equipment, or any design changes to the facility. Therefore,
based on the above, the proposed change does not represent a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the Primary Containment Leakage Rate
Testing Program in TS 5.5.12 for consistency with the requirements
of 10 CFR 50.55a(g)(4) for components classified as Code Class MC
and CC. The proposed change affects the frequency of visual
examinations that will be performed for the metallic and concrete
surfaces of containment and allows those examinations to be
performed during power operation as opposed to during a refueling
outage.
The proposed change does not involve a modification to the
physical configuration of the plants (i.e., no new equipment will be
installed), and does not revise the methods governing normal plant
operation. Also, the proposed change will not impose any new or
different requirements or introduce a new accident initiator,
accident precursor, or malfunction mechanism.
The proposed change also includes the removal of an item in TS
5.5.12 which was incorporated to establish deadlines for performing
the performance based Type A leakage tests in conjunction with
changing, on a one-time basis, the Type A test frequency. The
specified Unit 1 and Unit 2 Type A test have been completed. As
such, removal of this item is an administrative change.
As such, the proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the Primary Containment Leakage Rate
Testing Program in TS 5.5.12 for consistency with the requirements
of 10 CFR 50.55a(g)(4) for components classified as Code Class MC
and CC. The proposed change allows some of those examinations to be
performed during power operation as opposed to during a refueling
outage. As previously stated, the proposed change does not involve a
modification to the physical configuration of the plants and does
not revise the methods governing normal plant operation. As such,
the safety function of the containment as a fission product barrier,
will be maintained and is not adversely impacted by the proposed
change.
The proposed change also includes the removal of an item in TS
5.5.12 which was incorporated to establish deadlines for performing
the performance-based Type A leakage tests in conjunction with
changing, on a one-time basis, the Type A test frequency. The
specified Unit 1 and Unit 2 Type A test have been completed. As
such, removal of this item is an administrative change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336 Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: February 20, 2007.
Description of amendment request: The proposed amendment would
revise the Millstone Power Station, Unit No. 2 (MPS2) Technical
Specifications (TS) to eliminate Surveillance Requirement (SR) 4.5.2.e
which requires flow rate verification for each charging pump. Charging
pump flow is no longer relied upon for design basis mitigation at MPS2
and the charging pumps have been classified as non-risk significant in
the MPS2 Probabilistic Risk Assessment model. Therefore, the proposed
amendment is requesting to remove the charging pump flow verification
requirements currently located in the TS SR 4.5.2.e.
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 FSAR [Final Safety Analysis Report] Chapter 14 accident
analyses for MPS2 take no credit for the flow delivered by the
charging pumps. Additionally, the proposed change does not modify
any plant equipment or method of operation for any system, structure
or component required for safe operation of the facility or
mitigation of accidents assumed in the facility safety analyses. As
such, the proposed amendment does not increase the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not modify any plant equipment or
method of operation for any system, structure or component required
for safe operation of the facility or mitigation of accidents
assumed in the facility safety analyses. As such, no new failure
modes are introduced by the proposed change. Consequently, the
proposed amendment does not introduce any accident initiators or
malfunctions that would cause a new or different kind of accident.
Therefore, the proposed amendment does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The FSAR Chapter 14 accident analyses for MPS2 take no credit
for the charging pumps. The TS change does not involve a significant
reduction in a margin of safety because the proposed change does not
affect equipment design or operation, and there are no changes being
made to the technical specification required safety limits or safety
system settings. The proposed change does not affect any of the
assumptions used in the accident analysis, nor does it affect any
method of operation for equipment important to plant safety.
Therefore, the margin of safety is not impacted by the proposed
amendment.
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 Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Branch Chief: Harold K. Chernoff.
[[Page 68210]]
Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut.
Date of amendment request: July 2, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) 4.0.5 to reference the American
Society of Mechanical Engineers (ASME) Code for Operation and
Maintenance of Nuclear Power Plants (OM Code) instead of Section XI of
the ASME Boiler and Pressure Vessel Code. Specifically, the proposed
amendment would modify the inservice inspection (ISI) of ASME Code
Class 1, 2, and 3 components and inservice testing of ASME Code Class
1, 2, and 3 pumps and valves to reflect the requirements in the ASME OM
Code. In addition, the redundant requirement in TS 4.0.5 to maintain an
ISI program is being proposed for removal, based on duplicate
regulatory requirements set forth in Title 10 of the Code of Federal
Regulations (10 CFR) Section 50.55a.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not modify any plant equipment and does
not impact any failure modes that could lead to an accident.
Additionally, the proposed change has no effect on the consequence
of any analyzed accident since the change does not affect the
function of any equipment credited for accident mitigation. The
proposed change incorporates revisions to the ASME Code that result
in a net improvement in the measures for testing pumps and valves.
Removing from TS the duplicate requirement in the regulations to
maintain an ISI program in accordance with ASME codes and standards
does not impact any accident initiators or analyzed events or
mitigation of events. No reduction in previous commitments to 10 CFR
50.55a(g) are being proposed by this change.
Based on the discussion above, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or adversely affect methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. The proposed change does not
alter existing test criteria or frequencies. Additionally, there is
no change in the types or increases in the amounts of any effluent
that may be released off-site and there is no increase in individual
or cumulative occupational exposure. The proposed changes
incorporate revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves. Removal of
the duplicate TS requirement to maintain an ISI program will not
alter the commitment to the current ISI program requirements in 10
CFR 50.55a or any other TS requirements related to inservice
inspection.
Based on the discussion above, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises TS 4.0.5 regarding inservice testing
of ASME Code Class 1, 2, and 3 pumps and valves, for consistency
with the requirements of 10 CFR 50.55a(f)(4). The proposed change
incorporates an administrative clarification to the frequencies for
IST and incorporates revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves. No
setpoints or safety limit settings are being revised. The safety
function of the affected pumps and valves will continue to be
confirmed through inspection and testing. Removal of the ISI program
requirement from TS 4.0.5 does not remove the requirement from
regulations, and therefore, will not diminish the current station
approved programs and procedures that implement the regulatory
criteria of 10 CFR 50.55a(g) to maintain an acceptable ISI program
in accordance with the ASME Code.
Based on the discussion above, 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: Lillian M. Cuoco, Esquire, Senior Nuclear
Counsel, Dominion Nuclear Connecticut, Inc., Building 475, 5th Floor,
Rope Ferry Road, Waterford, CT 06141-5127.
NRC Branch Chief: Harold K. Chernoff.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Oconee Nuclear Station Independent Spent Fuel Storage Installation NRC
License No. SNM-2503, Docket No. 72-4, Oconee County, South Carolina
Date of amendment request: March 14, 2007.
Description of amendment request: The amendments would revise the
licenses to reflect the change in the name of the licensee from Duke
Power Company LLC to Duke Energy Carolinas, LLC. The proposed
amendments are a name change only. There is no change in the state of
incorporation, registered agent, registered office, rights or
liabilities of the company. Nor is there a change in the function of
the licensee or the way in which it does business.
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 amendments are for a name change only. The
amendments do not involve any change in the technical qualifications
of the licensee or the design, configuration, or operation of the
nuclear units. All Limiting Conditions for Operation, Limiting
Safety System Settings and Safety Limits specified in the Technical
Specifications remain unchanged. Also, the Physical Security Plans
and related plans, the Operator Training and Requalification
Programs, the Quality Assurance Programs, and the Emergency Plans
will not be materially changed by the proposed name change. The name
change amendments will not affect the executive oversight provided
by the Chief Nuclear Officer and his staff.
Therefore, the proposed amendments do not involve any increase
in the probability or consequences of an accident previously
analyzed.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
[[Page 68211]]
The proposed amendments do not involve any change in the design,
configuration, or operation of the nuclear plant. The current plant
design, design bases, and plant safety analysis will remain the
same.
The Limiting Conditions for Operations, Limiting Safety System
Settings and Safety Limits specified in the Technical Specifications
are not affected by the proposed changes. As such, the plant
conditions for which the design basis accident analyses were
performed remain valid.
The proposed amendments do not introduce a new mode of plant
operation or new accident precursors, do not involve any physical
alterations to plant configurations, or make changes to system
setpoints that could initiate a new or different kind of accident.
Therefore, the proposed amendments 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 amendments do not involve a change in the design,
configuration, or operation of the nuclear plants. The change does
not affect either the way in which the plant structures, systems,
and components perform their safety function or their design and
licensing bases.
Plant safety margins are established through Limiting Conditions
for Operation, Limiting Safety System Settings and Safety Limits
specified in the Technical Specifications. Because there is no
change to the physical design of the plant, there is no change to
any of these margins.
Therefore, the proposed amendments do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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: Evangelos C. Marinos.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: November 7, 2007.
Description of amendment request: The proposed amendment would
delete License Condition 2.F, which requires reporting of violations of
certain other requirements contained in Section 2.C of the license.
The NRC staff issued a ``Notice of Availability of Model
Application Concerning Elimination of Typical License Condition
Requiring Reporting of Violations of Section 2.C of Operating License
Using the Consolidated Line Item Improvement Process'' in the Federal
Register on November 4, 2005 (70 FR 67202). The notice referenced a
model safety evaluation, a model no significant hazards consideration
(NSHC) determination, and a model license amendment request published
in the Federal Register on August 29, 2005 (70 FR 51098). In its
application dated November 7, 2007, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee is presented below:
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 analysis adopted by the licensee
and, based upon this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendment involves NSHC.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2 (IP2), Westchester County, New York
Date of amendment request: October 24, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) requirements related to the
containment buffering agent used for pH control under post loss-of-
coolant accident (LOCA) conditions. Specifically, the proposal would
approve the use of sodium tetraborate (STB) as the buffering agent
instead of the currently approved compound, trisodium phosphate (TSP).
The reason for this change in buffering agents is to minimize the
potential for an adverse chemical interaction between the TSP and
certain insulation materials in the containment that could degrade flow
through the sump screens following certain design-basis accident
scenarios such as a LOCA.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response--No.
The proposed amendment does not involve a significant increase
in the probability of an accident previously evaluated because the
containment buffering agent is not an initiator of any analyzed
accident. The proposed change does not impact any failure modes that
could lead to an accident.
The proposed amendment does not involve a significant increase
in the consequences of an accident previously evaluated. The
buffering agent in containment is designed to buffer the acids
expected to be produced after a LOCA and is credited in the
radiological analysis for iodine retention. Utilizing STB as a
buffering agent ensures the post LOCA containment sump mixture will
have a pH >= 7.0. The proposed change of replacing TSP with STB
results in the radiological consequences remaining within the limits
of 10 CFR 50.67 as demonstrated by existing analyses of record.
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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response--No.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. STB is a passive component that is proposed to be used at
IP2 as a buffering agent to increase the pH of the initially acidic
post-LOCA containment water to a more neutral pH. Changing the
proposed buffering agent from TSP to STB does not constitute an
accident initiator or create a new or different
[[Page 68212]]
kind of accident previously analyzed. The proposed amendment does
not involve operation of any required systems, structures or
components in a manner or configuration different from those
previously recognized or evaluated. No new failure mechanisms will
be introduced by the changes being requested.
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 change involve a significant reduction in a
margin of safety?
Response--No.
The proposed amendment does not involve a significant reduction
in a margin of safety. The proposed amendment of changing the
buffering agent from TSP to STB results in equivalent control of
maintaining sump pH at 7.0 or greater, thereby controlling
containment atmosphere iodine and ensuring the radiological
consequences of a LOCA are within regulatory limits. The use of STB
also reduces the potential for exacerbating sump screen blockage due
to a chemical interaction between TSP and certain calcium sources
used in containment. This proposed amendment eliminates the
formation of calcium phosphate precipitate thereby reducing the
overall amount of precipitate that may be formed in a postulated
LOCA. The buffer change would minimize the potential chemical
effects and should enhance the ability of the emergency core cooling
system to perform the post-accident mitigating functions.
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: 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. 50-286, Indian Point
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York
Date of amendment request: October 24, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) requirements regarding the setpoint
and definition of the low-low level alarm on the Refueling Water
Storage Tank (RWST). Specifically, the proposal would revise the
setpoint of the low-low level alarm from a range of greater than or
equal to 10.5 ft and less than or equal to 12.5 ft to a range of
greater than or equal to 9.0 ft and less than or equal to 11.0 ft, and
revise the definition of the RWST ``low level alarm'' to ``low-low
level alarm.'' The reason for these changes is to ensure that adequate
water is supplied to the containment floor to eliminate the risk of
vortexing and/or draw down at the sump strainer modules following a
small-break loss-of-coolant accident (LOCA). The proposed changes are
being requested to support resolution of the pressurized-water reactor
sump performance issue involving debris accumulation, Generic Safety
Issue (GSI)-191.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the Technical Specifications are
consistent with the assumptions of all design basis accidents, as
they exist currently and as affected subsequent to the
implementation of the proposed amendment. The change in the RWST
low-low level alarm setpoint range has been demonstrated to be
within the safety margins for post-accident parameters and, in most
cases, actually beneficial to plant post-accident response
capability. The RWST is designed to respond to a variety of
accidents, and, for operation in Modes 1 through 4, it serves no
other purpose. Therefore, any adjustment of an intermediate level
setpoint cannot increase the probability of a design basis accident.
The change in the definition of the RWST ``low level alarm'' to
``low-low level alarm'' is editorial and therefore does not affect
the function of the alarm. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes represent a minor adjustment to an existing
setpoint range. The effect of the changes will be to assure
recirculation flow following a LOCA with consideration for sump
strainer installation, in response to GSI-191. However, the RWST
will continue to perform its function in essentially the same manner
that it has since original plant design. No changes in equipment
operation or procedural control will result from this amendment that
could possibly degrade the performance of the RWST or cause it to be
operated in a manner inconsistent with existing design basis
assumptions. The change in the definition of the RWST ``low level
alarm'' to ``low-low level alarm'' is editorial and therefore does
not affect the function of the alarm. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes improve the margin to safety, especially
with respect to post-accident temperature/pressure and dose
consequences during injection and, most importantly, pump
performance under postulated sump debris conditions during
recirculation. Significant margin is available to preclude air
ingestion in the ECCS [emergency core cooling system] pumps, and
sufficient time is available for the operators to perform the
switchover to recirculation. The change in the definition of the
RWST ``low level alarm'' to ``low-low level alarm'' is editorial and
therefore does not affect the function of the alarm. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
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
AmerGen Energy 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
[[Page 68213]]
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 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: July 19, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Sections 5.3.1/6.3.1, ``Unit Staff
Qualifications,'' for operator license applicants in accordance with
current industry standards for education and experience eligibility
requirements. The proposed amendment would permit changes to the unit
staff qualification education and experience eligibility requirements
for licensed operators. The proposal will bring Exelon Generation
Company, LLC (EGC) and AmerGen Energy Company, LLC (AmerGen) sites in
alignment with current industry practices and facilitate the
development of a pre-initial licensed operator training program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
Response: No.
Licensed operator qualification and training can have an
indirect impact on accidents previously evaluated. However, the NRC
considered this impact during the rulemaking process, and by
promulgation of the revised 10 CFR 55 rule, determined that this
impact remains acceptable when licensees have an accredited licensed
operator training program which is based on a systems approach to
training (SAT). The NRC has concluded in RIS [Regulatory Issue
Summary] 2001-01 and NUREG-1021 that standards and guidelines
applied by INPO [the Institute of Nuclear Power Operations] in their
accredited training programs are equivalent to those put forth by or
endorsed by the NRC. Therefore, maintaining an INPO accredited SAT
licensed operator training program is equivalent to maintaining an
NRC approved licensed operator training program which conforms with
applicable NRC Regulatory Guidelines or NRC endorsed industry
standards. The proposed changes conform to ACAD [air containment
atmosphere distribution] 00-003, Revision 1 licensed operator
education and experience eligibility requirements.
Based on the above, EGC and AmerGen conclude that the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Will operation of the facility in accordance with the
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
Response: No.
The proposed amendment involves changes to the licensed operator
training programs, which are administrative in nature. The EGC and
AmerGen licensed operator training programs have been accredited by
INPO and are based on SAT.
Based on the above discussion, EGC and AmerGen conclude that the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
The proposed TS changes are administrative in nature. The
proposed TS changes do not affect plant design, hardware, system
operation, or procedures for accident mitigation systems. The
proposed changes do not impact the performance or proficiency
requirements for licensed operators. As a result, the ability of the
plant to respond to and mitigate accidents is unchanged by the
proposed TS changes. Therefore, these changes do not involve a
significant reduction in a margin of safety.
Based on the above, EGC and AmerGen conclude that the proposed
changes do not involve a significant reduction in a margin of
safety.
Based on the above evaluation of the three criteria, EGC and
AmerGen conclude that the proposed amendment presents no significant
hazards consideration under the standards set forth in 10 CFR
50.92(c), and, accordingly, a finding of ``no significant hazards
consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
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 Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353,
Limerick Generating Station, Unit 1 and 2, Montgomery County,
Pennsylvania
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 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: August 8, 2007.
Description of amendment request: The proposed amendment replaces
references to Section XI of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code with references to the
ASME Code for Operation and Maintenance of Nuclear Power Plants (OM
Code) in the applicable technical specification (TS) section for the
Inservice Testing Program (IST) for the Exelon Generation Company, LLC,
and AmerGen Energy Company, LLC, (the licensees) plants that have
implemented industry Improved Technical Specifications. The proposed
changes are based on TS Task Force (TSTF) 479-A, Revision 0, ``Changes
to Reflect Revision of 10 CFR 50.55a,'' as modified by TSTF-497,
Revision 0, ``Limit Inservice Testing Program SR [Surveillance
Requirement] 3.0.2 Application to Frequencies of 2 Years or Less.'' In
addition, the proposed amendment adds a provision in the applicable TS
section to only apply the extension allowance of SR 3.0.2 to the
frequency table listed in the TS as part of the IST and to normal and
accelerated inservice testing frequencies of two years or less, as
applicable.
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 changes revise the applicable TS Section to conform
to the requirements of 10 CFR 50.55a, ``Codes and
[[Page 68214]]
standards,'' paragraph (f) regarding the inservice testing of pumps
and valves. The current TS reference the ASME Boiler and Pressure
Vessel Code, Section XI, requirements for the inservice testing of
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes
would reference the ASME OM Code as applicable, which is consistent
with 10 CFR 50 .55a, paragraph (f), ``Inservice testing
requirements.'' In addition, the proposed changes clarify that the
extension allowance of SR 3.0.2 only applies to the frequency table
listed in the TS, if applicable, as part of the Inservice Testing
Program and to normal and accelerated inservice testing frequencies
of two years or less. The definitions of the frequencies are not
changed by this license amendment request.
The proposed changes are administrative in nature, do not affect
any accident initiators, do not affect the ability to successfully
respond to previously evaluated accidents and do not affect
radiological assumptions used in the evaluations. Thus, the
probability or radiological consequences of any accident previously
evaluated are not increased.
Therefore, the proposed changes do not involve a significant
increase in 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 changes revise the applicable TS Section to conform
to the requirements of 10 CFR 50.55a(f) regarding the inservice
testing of pumps and valves. The current TS Section references the
ASME Boiler and Pressure Vessel Code, Section XI, requirements for
the inservice testing of ASME Code Class 1, 2, and 3 pumps and
valves. The proposed changes would reference the ASME OM Code as
applicable, which is consistent with 10 CFR 50.55a(f). In addition,
the proposed changes clarify that the extension allowance of SR
3.0.2 only applies to the frequency table listed in the TS, if
applicable, as part of the Inservice Testing Program and to normal
and accelerated inservice testing frequencies of two years or less.
The definitions of the frequencies are not changed by this license
amendment request.
The proposed changes to the applicable TS Section do not affect
the performance of any structure, system, or component credited with
mitigating any accident previously evaluated and do not introduce
any new modes of system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes revise the applicable TS Section for
Braidwood Station Units 1 and 2, Byron Station Units 1 and 2,
Dresden Nuclear Power Station Units 2 and 3, Limerick Generating
Station Units 1 and 2, Oyster Creek Generating Station, Peach Bottom
Atomic Power Station Units 2 and 3, Quad Cities Nuclear Power
Station Units 1 and 2, and Three Mile Island Unit 1 to conform to
the requirements of 10 CFR 50.55a(f) regarding the inservice testing
of pumps and valves.
The current TS Section references the ASME Boiler and Pressure
Vessel Code, Section XI, requirements for the inservice testing of
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes
would reference the ASME OM Code as applicable, which is consistent
with the 10 CFR 50.55a(f). In addition, the proposed changes clarify
that the extension allowance of SR 3.0.2 only applies to the
frequency table listed in the TS, if applicable, as part of the
Inservice Testing Program and to normal and accelerated inservice
testing frequencies of two years or less. The definitions of the
frequencies are not changed by this license amendment request.
The proposed changes do not modify the safety limits or
setpoints at which protective actions are initiated and do not
change the requirements governing operation or availability of
safety equipment assumed to operate to preserve the 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station
(QCNPS), Units 1 and 2, Rock Island County, Illinois
Date of application for amendment request: August 1, 2007.
Description of amendment request: The proposed amendment would
revise the technical specification (TS) allowable value (AV) for the
Reactor Protection System (RPS) Instrumentation Function 10, ``Turbine
Condenser Vacuum--Low,'' specified in TS Table 3.3.1.1-1, ``Reactor
Protection System Instrumentation.'' The proposed amendment also
revises the Channel Functional Test (CFT) and Channel Calibration (CC)
Surveillance Test Interval (STI) for DNPS TS Table 3.3.1.1-1, Function
10. As part of the DNPS STI revision, surveillance requirement (SR)
3.3.1.10, ``Channel Calibration,'' which is specific to the Turbine
Condenser Vacuum--Low instrument function, is deleted since it is no
longer applicable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Revision of Allowable Value
The proposed license amendment implements a revised AV for the
Turbine Condenser Vacuum--Low scram instrument function at DNPS,
Units 2 and 3 and QCNPS, Units 1 and 2.
The proposed changes to the DNPS and QCNPS Turbine Condenser
Vacuum--Low scram AV do not require modification [of] any system
interface or affect the probability of any event initiators at the
facilities. Overall RPS performance will remain within the bounds of
the previously performed accident analyses, since no hardware
changes are proposed.
There will be no degradation in the performance of, or an
increase in the number of challenges imposed on safety-related
equipment that are assumed to function during an accident situation.
The proposed changes will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in
the Updated Final Safety Analysis Report. The proposed changes are
consistent with safety analysis assumptions and resultant
consequences.
For these reasons, the proposed DNPS and QCNPS AV changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Relaxation of STIs (DNPS only)
The proposed license amendment implements a revised CFT and CC
STI for the Turbine Condenser Vacuum--Low scram instrument function
at DNPS Units 2 and 3. The proposed DNPS TS change to increase the
CFT STI for the Turbine Condenser Vacuum--Low scram instrument
function is based on an analytical method that has been reviewed and
approved by the NRC [Nuclear Regulatory Commission].
The proposed change to relax the CFT STI implements
recommendations from a generic evaluation that was developed by
General Electric (GE) and the Boiling Water Reactor Owners' Group
(BWROG), and subsequently approved by the NRC. This licensing
topical report (LTR) assessed the reliability of TS actuation
instrumentation and concluded that extending AOTS [allowed outage
times] and CFT STIs for test and repair activities would enhance
operational safety.
The proposed DNPS TS change to increase the CC STI for the
Turbine Condenser Vacuum--Low scram instrument function is based
upon a revised setpoint error analysis that provides revised AVs,
trip setpoints, and Expanded Tolerances (ETs) for the instrument.
These new AVs, trip setpoints,
[[Page 68215]]
and ETs establish increased design margin between the nominal trip
setpoint and the AV. This increased design margin, combined with
historical CC data, provides adequate assurance that the component
will remain operable when necessary for the prevention or mitigation
of accidents or transients.
The TS requirements that govern operability or routine testing
of plant instruments are not assumed to be initiators of any
analyzed event because these instruments are intended to prevent,
detect, or mitigate accidents. Therefore, these proposed STI changes
will not involve an increase in the probability of occurrence of an
accident previously evaluated. Additionally, these changes will not
increase the consequences of an accident previously evaluated
because the proposed changes do not involve any physical changes to
plant systems, structures or components (SSCs), or the manner in
which these SSCs are operated. These changes will not alter the
operation of equipment assumed to be available for the mitigation of
anticipated operational occurrences (AOOs) by the plant safety
analysis or licensing basis.
The proposed deletion of SR 3.3.1.10 is an administrative
change, since the SR will no longer be applicable to any instrument
function in DNPS TS Table 3.3.1-1. Therefore, the proposed deletion
of SR 3.3.1.10 will not impact the testing, calibration, and
inspection of RPS instrumentation that is necessary to assure that
the quality of the instrumentation is maintained, that facility
operation will be within safety limits, and that the limiting
conditions for operation will be met.
For these reasons, the proposed DNPS STI changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
In summary, the proposed license changes do not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to the DNPS and QCNPS Turbine Condenser
Vacuum--Low scram AV and the DNPS CFT and CC STIs do not affect the
design, functional performance, or operation of the facility.
Similarly, the proposed changes do not affect the design or
operation of any SSCs involved in the mitigation of any accidents,
nor do they affect the design or operation of any component in the
facilities such that new equipment failure modes are created.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed DNPS and QCNPS AV change does not affect the
acceptance criteria for any analyzed event, nor is there a change to
any Safety Analysis Limit. There will be no effect on the manner in
which safety limits, limiting safety system settings, or limiting
conditions for operation are determined nor will there be any effect
on those plant systems necessary to assure the accomplishment of
protection functions. All required design functions are maintained,
and the AVs, are consistent with NRC-approved methodology and
guidance for establishment of TS AVs.
The proposed AV changes do not affect the accident analyses that
assume operability of the instrument associated with the AV. The
Turbine Condenser Vacuum--Low scram function is credited in the Loss
of Main Condenser Vacuum AOO. The loss of main condenser AOO event
assumes that the main condenser is instantaneously lost while the
unit is operating at full power. This is classified as a moderate
frequency event and is described in the UFSAR [updated final safety
analysis report] as being bounded by the turbine trip with bypass
failure event.
The worst case for this AOO would occur if the loss of vacuum
were instantaneous. In this case, the loss of main condenser event
would be identical to the turbine trip with bypass failure event.
During a turbine trip with bypass failure event, the primary system
relief valves would remove the majority of the stored heat, while
the IC [isolation condenser] at DNPS and RCIC [reactor core
isolation cooling] at QCNPS would remove the remaining decay heat.
Slower losses of condenser vacuum would produce less severe AOOs,
since the turbine stop valves and bypass valves will still be
available prior to vacuum levels reaching the nominal trip setpoint
for the turbine trip and turbine bypass valve closure scram.
In that the proposed reduction of the Turbine Condenser Vacuum--
Low AV is based upon an AL [analytical limit] that is equal to the
nominal trip setpoint for the turbine trip, the resulting nominal
trip setpoint for the Turbine Condenser Vacuum--Low scram will still
be more conservative than the turbine trip setpoint. Therefore, the
sequence of events for the loss of main condenser AOO will still
result in a reactor scram prior to the turbine trip. Since the
proposed change to the Turbine Condenser Vacuum--Low AV will not
impact the limiting AOO analysis (i.e., the turbine trip with bypass
failure event), the proposed change does not reduce any margin of
safety.
Therefore, the proposed AV changes do not involve a significant
reduction in the margin of safety.
The proposed DNPS CFT STI change is based on an NRC-approved
generic analysis. This analysis concluded that the proposed CFT STI
change does not significantly affect the probability of failure or
availability of the affected instrumentation systems. Therefore, the
proposed DNPS CFT STI change does not affect the accident analyses
that assume operability of the instrument associated with the AV.
The proposed DNPS CC STI change is based on a revised setpoint
error analysis for the Turbine Condenser Vacuum--Low scram
instrument function that provides a revised AV, trip setpoint, and
Expanded Tolerance (ET) for the instrument. The new AV, trip
setpoint, and ET establish increased design margin between the
nominal trip setpoint and the AV. This increased design margin,
combined with historical CC data, provides adequate assurance that
the component will remain operable when necessary for the prevention
or mitigation of accidents or transients. Therefore, the proposed
DNPS CFT STI change does not affect the accident analyses that
assume operability of the instrument associated with the AV.
Therefore, the proposed changes to extend the DNPS CFT and CC
STIs do not involve a significant reduction in the margin of safety.
In summary, the proposed DNPS and QCNPS AV changes and DNPS STI
changes do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555. NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: October 9, 2007.
Description of amendment request: The proposed amendment would
change the technical specifications (TS) of Dresden Nuclear Power
Station (DNPS), Units 2 and 3, consistent with TS Task Force (TSTF)
Change Traveler TSTF-423 to the standard TSs boiling water reactor
plants, to allow, for some systems, entry into hot shutdown rather than
cold shutdown to repair equipment, if risk is assessed and managed
consistent with the program in place for complying with the
requirements of 10 CFR 50.65(a)(4). Changes proposed herein will be
made to the DNPS, Units 2 and 3, TSs for selected Required Action end
states providing this allowance.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on December 14, 2005 (70 FR 74037), on possible
license amendments adopting TSTF-423 using the NRC's consolidated line
item improvement process (CLIIP) for amending licensee's TSs, which
included a model safety evaluation (SE) and model no significant
hazards consideration (NSHC) determination. The NRC staff
[[Page 68216]]
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on March 26, 2006 (71 FR 14726), which included the resolution of
public comments on the model SE. The licensee affirmed the
applicability of the following NSHC determination in its application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
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 change to certain required end
states when the TS Completion Times for remaining in power operation
will be exceeded. Most of the requested technical specification (TS)
changes are to permit an end state of hot shutdown (Mode 3) rather
than an end state of cold shutdown (Mode 4) contained in the current
TS. The request was limited to: (1) Those end states where entry
into the shutdown mode is for a short interval, (2) entry is
initiated by inoperability of a single train of equipment or a
restriction on a plant operational parameter, unless otherwise
stated in the applicable technical specification, and (3) the
primary purpose is to correct the initiating condition and return to
power operation as soon as is practical. Risk insights from both the
qualitative and quantitative risk assessments were used in specific
TS assessments. Such assessments are documented in Section 6 of GE
NEDC-32988, Revision 2, ``Technical Justification to Support Risk
Informed Modification to Selected Required Action End States for BWR
Plants.'' They provide an integrated discussion of deterministic and
probabilistic issues, focusing on specific technical specifications,
which are used to support the proposed TS end state and associated
restrictions. The staff finds that the risk insights support the
conclusions of the specific TS assessments. Therefore, the
probability of an accident previously evaluated is not significantly
increased, if at all. The consequences of an accident after adopting
proposed TSTF-423, are no different than the consequences of an
accident prior to adopting TSTF-423. 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.
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).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded, i.e., entry into hot shutdown rather
than cold shutdown to repair equipment, 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 and the commitment by the licensee to adhere to the guidance
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0,
``Technical Specifications End States, NEDC-32988-A,'' 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.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The BWROG's risk assessment approach is
comprehensive and follows staff guidance as documented in RGs 1.174
and 1.177. In addition, the analyses show that the criteria of the
three-tiered approach for allowing TS changes are met. The risk
impact of the proposed TS changes was assessed following the three-
tiered approach recommended in RG 1.177. A risk assessment was
performed to justify the proposed TS changes. The net change to the
margin of safety is insignificant. Therefore, this change does not
involve a significant reduction in a margin of safety.
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.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: September 14, 2007.
Description of amendment request: Duane Arnold Energy Center
requests a proposed change to plant specific technical specifications
(TS) 3.3.2.1, ``Control Rod Block Instrumentation,'' to allow the use
of the improved Banked Position Withdrawal Sequence (BPWS) during
shutdowns in accordance with NEDO-33091-A, Revision 2, ``Improved BPWS
Control Rod Insertion Process,'' dated July 2004. The proposed changes
are consistent with Nuclear Regulatory Commission (NRC)-approved
Industry Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-476, Revision 1, ``Improved BPWS
Control Rod Insertion Process (NEDO-33091).''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no-significant-hazards-consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed changes modify the TS to allow the use of the improved
banked position withdrawal sequence (BPWS) during shutdowns if the
conditions of NEDO-33091-A, Revision 2, ``Improved BPWS Control Rod
Insertion Process,'' July 2004, have been satisfied. The staff finds
that the licensee's justifications to support the specific TS changes
are consistent with the approved topical report and TSTF-476, Revision
1. Since the change only involves changes in control rod sequencing,
the probability of an accident previously evaluated is not
significantly increased, if at all. The consequences of an accident
after adopting TSTF-476 are no different than the consequences of an
accident prior to adopting TSTF-476. Therefore, the consequences of an
accident previously evaluated are not significantly affected by this
change. 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 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 control rod drop accident (CRDA) is the
design basis accident for the subject TS changes. 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, TSTF-476, Revision 1, incorporates the
improved BPWS, previously approved in NEDO-33091-A, into the improved
TS. The control rod drop accident (CRDA) is the design basis accident
for the subject TS changes. In order to minimize the impact of a CRDA,
the BPWS process was developed to minimize control rod reactivity worth
for BWR plants. The proposed improved BPWS further simplifies the
control rod insertion process, and in order to evaluate it, the
[[Page 68217]]
staff followed the guidelines of Standard Review Plan Section 15.4.9,
and referred to General Design Criterion 28 of Appendix A to 10 CFR
Part 50 as its regulatory requirement. The TSTF stated the improved
BPWS provides the following benefits: (1) Allows the plant to reach the
all-rods-in condition prior to significant reactor cool down, which
reduces the potential for re-criticality as the reactor cools down; (2)
reduces the potential for an operator reactivity control error by
reducing the total number of control rod manipulations; (3) minimizes
the need for manual scrams during plant shutdowns, resulting in less
wear on control rod drive (CRD) system components and CRD mechanisms;
and, (4) eliminates unnecessary control rod manipulations at low power,
resulting in less wear on reactor manual control and CRD system
components. The addition of procedural requirements and verifications
specified in NEDO-33091-A, along with the proper use of the BPWS will
prevent a control rod drop accident (CRDA) from occurring while power
is below the low power setpoint (LPSP). The net change to the margin of
safety is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
Based upon the above discussion of the amendment request, the
requested change does not involve a significant hazards consideration.
Attorney for licensee: Marjan Mashhadi, Florida Power & Light
Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.
NRC Acting Branch Chief: Clifford G. Munson.
FPL Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach
Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County,
Wisconsin
Date of amendment request: October 12, 2007.
Description of amendment request: FPL Energy Point Beach, LLC
(FPLE-PB) proposes to revise Technical Specification (TS) 5.5.1 5
``Containment Leakage Rate Testing Program,'' for Units 1 and 2. The
proposed change would allow a one-time interval extension of no more
than 5 years for the Type A, Integrated Leakage Rate Test (ILRT).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This license amendment proposes to revise the Technical
Specifications (TS) to allow for the one-time extension of the
containment integrated leakage rate test interval from 10 to 15
years. The containment vessel function is to mitigate consequences
of an accident. There are no design basis accidents initiated by a
failure of the containment leakage mitigation function. The
extension of the containment integrated leakage rate test interval
will not create an adverse interaction with other systems that could
result in initiation of a design basis accident. Therefore, the
probability of occurrence of an accident previously evaluated is not
significantly increased.
The potential consequences of the proposed change have been
quantified by analyzing the changes in risk that would result from
extending the containment integrated leakage rate test interval from
10 to 15 years. The increase in risk in terms of person-rem per year
within 50 miles resulting from design basis accidents was estimated
to be of a magnitude that NUREG-1493 indicates is very small. FPLE-
PB has also analyzed the increase in risk in terms of the frequency
of large early releases from accidents. The increase in the large
early release frequency resulting from the proposed extension was
determined to be within the guidelines published in RG 1.I74.
Additionally, the proposed change maintains defense-in-depth by
preserving a reasonable balance among prevention of core damage,
prevention of containment failure, and consequence mitigation. FPLE-
PB has determined that the increase in conditional containment
failure probability from reducing the containment integrated leakage
rate test frequency from one test per 10 years to one test per 15
years would be small.
Continued containment integrity is also assured by the history
of successful containment integrated leakage rate tests, and the
established programs for local leakage rate testing and IWE
inservice inspections which are not affected by the proposed change.
Therefore, the probability of occurrence or the consequences of an
accident previously analyzed are not significantly increased.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to extend the containment integrated leakage
rate test interval from 10 to 15 years does not create any new or
different accident initiators or precursors. The length of the
containment integrated leakage rate test interval does not affect
the manner in which any accident begins. The proposed change does
not create any new failure modes for the containment and does not
affect the interaction between the containment and any other system.
Thus, the proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The risk-based margins of safety associated with the containment
integrated leakage rate test are those associated with the estimated
person-rem per year, the large early release frequency and the
conditional containment failure probability. FPLE-PB has quantified
the potential effect of the proposed change on these parameters and
determined that the effect is not significant. The non-risk-based
margins of safety associated with the containment integrated leakage
rate test are those involved with its structural integrity and leak
tightness. The proposed change to extend the containment integrated
leakage rate test interval from 10 to 15 years does not adversely
affect either of these attributes. The proposed change only affects
the frequency at which these attributes are verified. Therefore, the
proposed change does not involve a significant reduction in 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. Antonio Fernandez, Senior Attorney, FPL
Energy, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Acting Branch Chief: Cliff Munson.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: October 22, 2007.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) 3.6.3.1, Primary Containment
Hydrogen Recombiners, and references to the hydrogen and oxygen
monitors in TS 3.3.3.1, Post Accident Monitoring (PAM) Instrumentation.
The proposed TS changes support implementation of the revisions to
Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44,
``Combustible gas control for nuclear power reactors,'' that became
effective on October 16, 2003. These changes are consistent with
Nuclear Regulatory Commission (NRC)-approved Revision 1 to TS Task
Force (TSTF) Change Traveler, TSTF-447, ``Elimination of Hydrogen
Recombiners and Change to Hydrogen and Oxygen Monitors.'' The
availability of this TS improvement was announced in the Federal
Register on September 25, 2003 (68 FR 55416) as part of the
consolidated line item improvement process. The licensee affirmed the
applicability of the model no significant hazards consideration
determination in its application.
[[Page 68218]]
The proposed amendment would also relocate, from the Renewed
Facility Operating License to the NMP2 Updated Safety Analysis Report,
License paragraph 2.C.(11a), Additional Condition 3, which requires
establishing containment hydrogen monitoring within 90 minutes of
initiating emergency core cooling following a loss-of-coolant accident.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration adopted by the licensee is
presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-of-
coolant accident (LOCA) hydrogen release, and eliminates requirements
for hydrogen control systems to mitigate such a release. The
installation of hydrogen recombiners and/or vent and purge systems
required by 10 CFR 50.44(b)(3) was intended to address the limited
quantity and rate of hydrogen generation that was postulated from a
design-basis LOCA. The Commission has found that this hydrogen release
is not risk-significant because the design-basis LOCA hydrogen release
does not contribute to the conditional probability of a large release
up to approximately 24 hours after the onset of core damage. In
addition, these systems were ineffective at mitigating hydrogen
releases from risk-significant accident sequences that could threaten
containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen [and oxygen] monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10 CFR
50.2. RG 1.97 Category 1 is intended for key variables that most
directly indicate the accomplishment of a safety function for design-
basis accident events. The hydrogen [and oxygen] monitors no longer
meet the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as defined
in RG 1.97, is an appropriate categorization for the hydrogen monitors
because the monitors are required to diagnose the course of beyond
design-basis accidents. [Also, as part of the rulemaking to revise 10
CFR 50.44, the Commission found that Category 2, as defined in RG 1.97,
is an appropriate categorization for the oxygen monitors, because the
monitors are required to verify the status of the inert containment.]
The regulatory requirements for the hydrogen [and oxygen] monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite releases
of radioactivity, and establishing protective action recommendations to
be communicated to offsite authorities. Classification of the hydrogen
monitors as Category 3, [classification of the oxygen monitors as
Category 2] and removal of the hydrogen [and oxygen] monitors from TS
will not prevent an accident management strategy through the use of the
SAMGs [severe accident management guidelines], the emergency plan (EP),
the emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner requirements
and relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident from Any [Accident] Previously
Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements, including
removal of these requirements from TS, will not result in any failure
mode not previously analyzed. The hydrogen recombiner and hydrogen [and
oxygen] monitor equipment was intended to mitigate a design-basis
hydrogen release. The hydrogen recombiner and hydrogen [and oxygen]
monitor equipment are not considered accident precursors, nor does
their existence or elimination have any adverse impact on the pre-
accident state of the reactor core or post accident confinement of
radionuclides within the containment building.
Therefore, 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 [a] Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements, including
removal of these requirements from TS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results in
a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this hydrogen
release is not risk-significant because the design-basis LOCA hydrogen
release does not contribute to the conditional probability of a large
release up to approximately 24 hours after the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can be
adequately met without reliance on safety-related hydrogen monitors.
[Category 2 oxygen monitors are adequate to verify the status of an
inerted containment.]
Therefore, this change does not involve a significant reduction in
[a] margin of safety. [The intent of the requirements established as a
result of the TMI, Unit 2 accident can be adequately met without
reliance on safety-related oxygen monitors.] Removal of hydrogen [and
oxygen] monitoring from TS will not result in a significant reduction
in their functionality, reliability, and availability.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendments involves no significant hazards
consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of amendment request: October 17, 2007.
Description of amendment request: The proposed amendment would
allow a one-time revision to the requirements for fuel decay time prior
to commencing movement of irradiated fuel in the
[[Page 68219]]
reactor pressure vessel (RPV). Currently, Technical Specification (TS)
3/4.9.3, ``Decay Time'' requires that: (a) The reactor has been
subcritical for at least 100 hours prior to movement of irradiated fuel
in the RPV between October 15th through May 15th; and (b) the reactor
has been subcritical for at least 168 hours prior to movement of
irradiated fuel in the RPV between May 16th and October 14th. The
calendar approach is based on average river water temperature which is
cooler in the fall through spring months. The proposed amendment would
revise TS 3/4.9.3 to allow fuel movement to commence at 86 hours after
the reactor is subcritical. The proposed change would only be
applicable to Salem Nuclear Generating Station, Unit No. 2 refueling
outage 2R16, which is scheduled to commence on March 4, 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 license amendment would allow fuel assemblies to be
removed from the reactor core and be stored in the Spent Fuel Pool
[SFP] in less time after subcriticality than currently allowed by
the TSs. Decreasing the decay time of the fuel affects the
radionuclide make-up of the fuel to be offloaded as well as the
amount of decay heat that is present from the fuel at the time of
offload. The accident previously evaluated that is associated with
the proposed license amendment is the fuel handling accident [FHA].
Allowing the fuel to be offloaded in less time after subcriticality
using actual heat loads does not impact the manner in which the fuel
is offloaded. The accident initiator is the dropping of the fuel
assembly. Since earlier offload does not affect fuel handling, there
is no increase in the probability of occurrence of a [FHA]. The time
frame in which the fuel assemblies are moved has been evaluated
against the [Title 10 of the Code of Federal Regulations (10 CFR)
Section 50.67] dose limits for members of the public, licensee
personnel and control room. Additionally, the guidance provided in
[Regulatory Guide (RG)] 1.183 was used for the selective application
of Alternative Source Term. All dose limits are met with the reduced
core offload times; and significant margin is maintained, as the
minimum decay time prior to movement of fuel for the FHA analysis is
24 hours.
Therefore, the proposed license amendment does not significantly
increase the probability [ ] or the consequences of accidents
previously evaluated.
2. [Does the change] [c]reate the possibility of a new or
different kind of accident from any accident previously evaluated[?]
Response: No.
The proposed license amendment would allow core offload to occur
in less time after subcriticality which affects the radionuclide
makeup of the fuel to be offloaded as well as the amount of decay
heat that is present from the fuel at the time of offload. The
radionuclide makeup of the fuel assemblies and the amount of decay
heat produced by the fuel assemblies do not currently initiate any
accident. A change in the radionuclide makeup of the fuel at the
time of core offload or an increase in the decay heat produced by
the fuel being offloaded will not cause the initiation of any
accident. The accident previously evaluated that is associated with
fuel movement is the [FHA]; no new accidents are introduced. There
is no change to the manner in which fuel is being handled or in the
equipment used to offload or store the fuel. The effects of the
additional decay heat load have been analyzed. The analysis
demonstrates that the existing [SFP] cooling system and associated
systems under worst-case circumstances would maintain licensing
limits and the integrity of the [SFP].
Therefore, the proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The margin of safety pertinent to the proposed changes is the
dose consequences resulting from a [FHA]. The shorter decay time
prior to fuel movement has been evaluated against 10 CFR 50.67 and
all limits continue to be met. All dose limits are met with the
reduced core offload times; and significant margin is maintained, as
the minimum decay time prior to movement of fuel for the FHA
analysis is 24 hours. Decay heat-up calculations performed prior to
the refueling outage as part of the IDHM [Integrated Decay Heat
Management] program ensure that planned spent fuel transfer to the
SFP will not result in maximum SFP temperature exceeding the design
basis limit of 149[deg]F (with both heat exchangers available) or
180[deg]F (with one heat exchanger alternating between the two
pools). As stated above, the changes in radionuclide makeup and
additional heat load do not impact any safety settings and do not
cause any safety limit to not be met. In addition, the integrity of
the [SFP] is maintained.
The time frame in which the fuel assemblies are moved has been
evaluated against the 10 CFR 50.67 dose limits for members of the
public, licensee personnel and control room. Additionally, the
guidance provided in [RG] 1.183 was used. Calculations performed
conclude that expected dose limits following a [FHA] are met with
the proposed decay time prior to commencing fuel movement.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Unit Nos. 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 27, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) to establish more effective
and appropriate action, surveillance, and administrative requirements
related to ensuring the habitability of the control room envelope (CRE)
in accordance with Nuclear Regulatory Commission (NRC)-approved
Technical Specification Task Force (TSTF) Standard Technical
Specification change traveler TSTF-448, Revision 3, ``Control Room
Habitability.'' Specifically, the proposed amendment would modify TS
3.7.7, ``Control Room Emergency Ventilation System,'' and TS Section 6,
``Administrative Controls.'' The NRC staff issued a ``Notice of
Availability of Technical Specification Improvement to Modify
Requirements Regarding Control Room Envelope Habitability Using the
Consolidated Line Item Improvement Process associated with TSTF-448,
Revision 3, in the Federal Register on January 17, 2007 (72 FR 2022).
The notice included a model safety evaluation, a model no significant
hazards consideration (NSHC) determination, and a model license
amendment request. In its application dated October 27, 2007, Tennessee
Valley Authority (the licensee) affirmed the applicability of the model
NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration adopted by the licensee is
presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability
[[Page 68220]]
of structures, systems, and components to perform their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change revises the TS
for the CRE emergency ventilation system, which is a mitigation
system designed to minimize unfiltered air leakage into the CRE and
to filter the CRE atmosphere to protect the CRE occupants in the
event of accidents previously analyzed. An important part of the CRE
emergency ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed 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 impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, 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.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendments involves no significant hazards
consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
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., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: February 16, 2007.
Brief description of amendment request: The proposed amendment
would revise Technical Specification 3/4.4.3, ``Reactor Coolant System,
Relief Valves'' to modify the method of testing the pressurizer Power
Operated Relief Valves (PORVs). Specifically the requirement for bench
testing the valves is changed to accommodate testing of the PORVs while
installed in the plant. The change is requested due to the installation
of new PORVs that are welded to the piping rather than bolted into the
system.
Date of publication of individual notice in Federal Register:
November 19, 2007.
Expiration date of individual notice: December 19, 2007 (public
comment), January 18, 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].
[[Page 68221]]
Duke Power Company LLC, et. al., Docket No. 50-414, Catawba Nuclear
Station, Unit 2, York County, South Carolina
Date of application for amendments: April 30, 2007.
Brief description of amendments: The amendment revised Technical
Specification (TS) 5.5.9, ``team Generator (SG) Tube Surveillance
Program,'' regarding the required SG inspection scope for Catawba Unit
2 during the End of Cycle 15 Refueling Outage and Operating Cycle 16.
The changes modified the tube repair criteria for portions of the SG
tubes within the hot leg tubesheet region of the SGs.
Date of issuance: October 31, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 233.
Renewed Facility Operating License No. NPF-52: Amendments revised
the licenses and the technical specifications.
Date of initial notice in Federal Register: August 13, 2007 (72 FR
45272).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et. al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: March 29, 2007, as supplemented
September 7, 2007, October 9 and October 12, 2007.
Brief description of amendments: The amendments revised the Catawba
1 and 2, Technical Specifications 3.5.2.8, and authorized changes to
the updated final safety analysis report concerning modifications to
the emergency core cooling system sump.
Date of issuance: November 8, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 238, 234.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: August 13, 2007 (72 FR
45274). The supplements dated September 7, 2007, October 9, and October
12, 2007, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 8, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: November 16, 2006, supplemented
May 9 and August 28, 2007.
Brief description of amendments: The amendments authorized revision
of the Updated Final Safety Analysis Report to describe the flood
protection measures for the auxiliary building.
Date of Issuance: November 14, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days after completion of the flood protection measures for
the auxiliary building.
Amendment Nos.: 357, 359, and 358.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
151). The supplements dated May 9 and August 28, 2007, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 14, 2007.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: July 16, 2007, as supplemented by letter
dated August 7, 2007.
Brief description of amendment: The proposed amendment revised the
facility operating license (FOL), Paragraph 2.C, and technical
specifications (TS) 3.7.2 and TS 5.5 for River Bend Station, Unit 1.
Date of issuance: November 16, 2007.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 154.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51857). The supplement dated August 7, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register on September 11, 2007 (72 FR 51857).
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated November 16, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendment: November 27, 2006, as
supplemented by letter dated August 24, 2007.
Brief description of amendment: This amendment revises multiple TSs
relating to testing of the Emergency Diesel Generators (EDGs).
Specifically, the changes eliminate various accelerated testing
requirements, eliminate the EDG test schedule table based on failure
rates, relax acceptance criteria associated with the ``fast start'' and
load rejection tests and eliminate the EDG failure report.
Date of issuance: November 6, 2007.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 189 and 150.
Facility Operating License Nos. NPF-39 and NPF-85: This amendment
revised the license and Technical Specifications.
Date of initial notice in Federal Register: July 31, 2007 (72 FR
41784).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 6, 2007.
No significant hazards consideration comments received: No.
FPL Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach
Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County,
Wisconsin
Date of application for amendments: June 29, 2007.
Brief description of amendments: The amendments would modify the
Technical Specifications (TSs) 3.7.2, by removing the specific
isolation time for the main steam isolation valves from the
[[Page 68222]]
associated TS surveillance requirements and by replacing it with the
requirement to verify the valve isolation time is within limits.
Date of issuance: November 16, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 230, 235.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Technical Specifications/License.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51865).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 16, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: July 9, 2007.
Brief description of amendment: The amendment revised the Technical
Specifications by removing the Table of Contents.
Date of issuance: November 8, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 152.
Facility Operating License No. DPR-22.
Amendment revised the Technical Specifications. Date of initial
notice in Federal Register: August 14, 2007 (72 FR 45459).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 8, 2007.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, SalemNuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: August 15, 2007, as
supplemented on September 6, 2007.
Brief description of amendments: The amendments revise the
licensing basis, as described in Appendix 3A of the Salem Updated Final
Safety Analysis Report (UFSAR), regarding the method of calculating the
net positive suction head available for the emergency core cooling
system and containment heat removal system pumps. These changes to the
Salem licensing basis relate to issues associated with Generic Letter
2004-02, ``Potential Impact of Debris Blockage on Emergency
Recirculation During Design Basis Accidents at Pressurized-Water
Reactors.''
Date of issuance: November 15, 2007.
Effective date: As of the date of issuance, to be implemented by
December 31, 2007.
Amendment Nos.: 285 and 268.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revise the UFSAR.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51866). The letter dated September 6, 2007, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the application beyond
the scope of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 15, 2007.
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
[[Page 68223]]
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 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
[[Page 68224]]
download the Workplace Forms Viewer\TM\ to access the Electronic
Information Exchange (EIE), a component of the E-Filing system. The
Workplace Forms Viewer\TM\ 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.
Virginia Electric and Power Company, et. al., Docket Nos. 50-280 and
50-281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: October 22, 2007, as
supplemented November 2 and November 9, 2007.
Brief Description of amendments: This amendment adds a new license
condition, P.(3), to license Nos. DPR-32 and DPR-37, which authorize
the licensee to modify the GOTHIC code as described in the Updated
Final Safety Analysis Report (UFSAR) and update the UFSAR as required
by 10 CFR 50.71(e).
Date of issuance: November 15, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 256, 255.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments revise the licenses.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. The notice provided an opportunity to submit
comments (by November 13, 2007) on the Commission(s proposed NSHC
determination. No comments have been received. The notice also provided
an opportunity to request a hearing (by December 31, 2007), but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment. The
Commission's related evaluation of the amendment, finding of exigent
circumstances, state consultation, and final NSHC determination are
contained in a safety evaluation dated November 15, 2007.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq.
NRC Branch Chief: Evangelos C. Marinos.
Dated at Rockville, Maryland, this 23rd day of November, 2007.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-23225 Filed 12-3-07; 8:45 am]
BILLING CODE 7590-01-P