[Federal Register Volume 71, Number 215 (Tuesday, November 7, 2006)]
[Notices]
[Pages 65139-65148]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-18595]
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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 October 13, 2006, to October 26, 2006. The
last biweekly notice was published on October 24, 2006 (71 FR 62306).
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, 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. 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 persons 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
[[Page 65140]]
request for a hearing or petition for leave to intervene is filed
within 60 days, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed 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; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
For further details with respect to this 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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: September 28, 2006.
Description of amendments request: The proposed amendments would
revise certain Technical Specification (TS) requirements for mode
change limitations in Limiting Condition for Operation 3.0.4 and
Surveillance Requirement 3.0.4. This request is consistent with NRC-
approved Industry/Technical Specification Task Force (TSTF) Traveler
number TSTF-359, Revision 9, ``Increase Flexibility in Mode
Restraints.'' In addition, the proposed amendments would correct TS
Example 1.4-1, ``Surveillance Requirements,'' to accurately reflect the
changes made by TSTF-359, which is consistent with NRC-approved TSTF-
485, Revision 0.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Section 1.4, Frequency, ``Example
1.4-1,'' to be consistent with Surveillance Requirement (SR) 3.0.4
and Limiting Condition for Operation (LCO) 3.0.4. This change is
considered administrative in that it modifies the example to
demonstrate the proper application of SR 3.0.4 and LCO 3.0.4. The
requirements of SR 3.0.4 and LCO 3.0.4 are clear and are clearly
explained in the associated Bases. As a result, modifying the
example will not result in a change in usage
[[Page 65141]]
of the Technical Specifications (TS). The proposed change does not
adversely affect accident initiators or precursors, the ability of
structures, systems, and components (SSCs) to perform their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Therefore, this change is considered
administrative and will have no effect on the probability or
consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the change does not impose any new or
different requirements or eliminate any existing requirements. The
change does not alter assumptions made in the safety analysis. The
proposed change is consistent with the safety analysis assumptions
and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is administrative and will have no effect on
the application of the Technical Specification requirements.
Therefore, the margin of safety provided by the Technical
Specification requirements is unchanged.
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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: David Terao.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2 New London County, Connecticut
Date of amendment request: March 17, 2006.
Description of amendment request: The proposed amendment would
revise Millstone Power Station, Unit No. 2 Technical Specification (TS)
3.4.4 to replace the existing maximum and minimum pressurizer water
volume and water level limits with a maximum water level limit. The
associated TS bases will be updated to address the proposed change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not change the accident analysis of
record, maintains the current maximum operating pressurizer level at
its present value, 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 consequences of any analyzed
accident since the change does not affect the function of any equipment
credited for accident mitigation. Therefore, the proposed amendment
does not increase the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Since the proposed change does not modify any plant equipment and
there is no impact on the capability of existing equipment to perform
its intended functions and no new failure modes are introduced by the
proposed change, the proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety?
The proposed change maintains the current maximum operating
pressurizer level at its present value, and the acceptance criterion
for the maximum pressurizer level is unchanged. Since there are no
changes, the proposed change does not involve a reduction in a margin
of safety.
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.
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: October 16, 2006.
Description of amendment request: The proposed change will add an
NRC previously approved topical report to the analytical methods
referenced in Technical Specification (TS) Section 5.6.5, ``Core
Operating Limits Report (COLR).'' The current method of performing the
loss-of-coolant accident (LOCA ) analyses will be replaced by an
updated method described in AREVA NP (formerly known as Framatome or
Siemens) topical report, ``EXEM BWR-2000 [Boiling-Water Reactor--2000]
ECCS [Emergency Core Cooling System] Evaluation Model.''
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.
Core operating limits are established each operating cycle in
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core
Operating Limits Report (COLR)''. These core operating limits ensure
that the fuel design limits are not exceeded during any conditions
of normal operation or in the event of any Anticipated Operational
Occurrence (AOO). In addition, the Average Planar Linear Heat
Generation Rate (APLHGR) operating limits imposed by Technical
Specification 3.2.1 also ensure that the peak cladding temperature
(PCT) during the postulated design basis LOCA does not exceed the
2200 [deg]F limit specified in 10 CFR 50.46. The APLHGR is a measure
of the average linear heat generation rate of all the fuel rods in a
fuel assembly at any axial location.
The methods used to determine the operating limits are those
previously found acceptable by the NRC and listed in TS section
5.6.5.b. A change to TS section 5.6.5.b is requested to include an
updated LOCA analysis method, EXEM BWR-2000. The updated method will
be used to determine the APLHGR operating limits imposed by
Technical Specification 3.2.1. EXEM BWR-2000 has been reviewed and
approved by the NRC and is applicable to the RBS [River Bend
Station] plant design and the AREVA NP fuel being used at RBS. The
application of the LOCA analytical model will continue to ensure
that the APLHGR
[[Page 65142]]
operating limits are established to protect the fuel cladding
integrity during normal operation, AOOs, and the design basis LOCA.
The requested TS changes concern the use of analytical methods
and do not involve any plant modifications or operational changes
that could affect any postulated accident precursors or accident
mitigation systems and do not introduce any new accident initiation
mechanisms.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS amendment will not change the design function,
reliability, performance, or operation of any plant systems,
components, or structures. It does not create the possibility of a
new failure mechanism, malfunction, or accident initiators not
considered in the design and licensing bases. Plant operation will
continue to be within the core operating limits that are established
using NRC approved methods that are applicable to the RBS design and
the RBS fuel.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The ECCS performance analysis methods are used to establish the
APLHGR limits required by Technical Specification 3.2.1. The APLHGR
limits are specified in the COLR and are the result of fuel design,
design basis accident (DBA), and transient analyses. Limits on the
APLHGR are specified to ensure that the fuel design limits are not
exceeded during anticipated operational occurrences (AOOs) and that
the peak cladding temperature (PCT) during the postulated design
basis LOCA does not exceed the 2200 [deg]F limit specified in 10 CFR
50.46.
The EXEM BWR-2000 evaluation model is an updated LOCA analytical
method that has been approved by the NRC and is applicable to the
RBS plant design and the fuel being used at RBS. A RBS plant
specific ECCS performance analysis has been performed with the EXEM
BWR-2000 evaluation model. This evaluation concluded that the
resulting PCT still afforded adequate margin to the 2200 [deg]F
limit of 10 CFR 50.46.
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: Mark Wetterhahn, Esq., Winston & Strawn LLP,
1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: David Terao.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of amendment request: September 25, 2006.
Description of amendment request: The amendment proposes revisions
to the Technical Specifications that are editorial in nature and
consist of typographical corrections, update of references, and
deletion of obsolete notes.
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 are editorial in nature and have no affect
on accident scenarios previously evaluated. The proposed changes
include typographical corrections, consistent with the current
version of the Standard Technical Specifications (NUREG 1431,
Revision 3); updated references, consistent with the current version
of the Entergy Quality Assurance Program Manual (Revision 13); and
deletion of notes that provided one-time allowances or are otherwise
now obsolete. The proposed changes do not affect initiating events
for accidents previously evaluated and do not affect or modify
plants systems or procedures used to mitigate the progression or
outcome of those accident scenarios.
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 do not involve the installation of new
plant equipment or modification of existing plant equipment. No
system or component setpoints are being changed and there are no
changes being proposed for the way that the plant is operated. There
are no new accident initiators or equipment failure modes resulting
from the proposed changes. The proposed changes are editorial in
nature, consisting of typographical corrections, reference updates,
and deletion of obsolete notes.
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 are editorial in nature and do not change
setpoints or limiting parameters specified in the plant Technical
Specifications.
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. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Richard J. Laufer.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: August 31, 2006.
Description of amendment request: Entergy Operations, Inc.,
proposes to relocate Technical Specification (TS) 3.8.7 requirements
associated with 120 Volt Inverter Y-28 and TS 3.8.9 requirements
associated with 120 VAC electrical power distribution subsystem panel
C-540 to the Technical Requirements Manual (TRM).
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 does not physically alter any plant
structures, systems, or components and does not affect or create new
accident initiators or precursors. The loss of Y-28, in itself, has
no significant impact on station operation because its associated
instrument panel, C-540, remains energized from an Emergency Diesel
Generator (EDG) backed vital AC source. A potential loss of vital
instrument panel C-540 does not prevent the fulfillment of a safety
function and does not cause Emergency Safeguard Features (ESF)
systems actuations that could render multiple ESF-related trains
incapable of performing their intended safety function. Therefore,
there is no effect on probability of accidents previously evaluated.
The proposed change relocates operability requirements for Y-28
and C-540 to the TRM. The TRM is part of the Safety Analysis Report
(SAR) and is controlled under 10 CFR 50.59. In addition, TS-related
components powered by C-540 continue to be governed by other TSs
that limit the time in which the
[[Page 65143]]
components can be out of service or provide compensatory measures
during the out-of-service period.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not physically alter any structures,
systems, or components, and does not affect or create new accident
initiators or precursors. The accident analysis assumptions and
results are unchanged. No new failures or interactions have been
created. In addition, the proposed change does not introduce new
failure modes or mechanisms associated with plant operation and will
not create a new accident type.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The applicable margin of safety is the period of time that
equipment important to safety is inoperable. There is no increase in
risk that is a result of the proposed change because (1) affected
non-TS components are not safety significant, (2) compensatory
measures are procedurally established for those components governed
by other regulation (i.e., 10 CFR [Part] 50, Appendix R), and (3)
TS-related component out-of-service time or related compensatory
actions are governed by other existing TSs. The proposed change does
not affect any safety limits, other operational parameters, or
setpoints in the TS, nor does it affect any margins assumed in the
accident analyses. In addition, Y-28 and C-540 operability
requirements will be relocated to the TRM, which is part of the
Safety Analysis Report (SAR) and controlled by 10 CFR 50.59.
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendments request: June 30, 2006.
Description of amendments request: The amendments would relocate
the movable incore detectors and radioactive gaseous effluent oxygen
monitoring instrumentation from the Technical Specifications to the
Updated Final Safety Analysis Report (UFSAR).
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 change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would relocate Technical Specification
(TS) 3/4.3.3.2, ``Movable Incore Detectors,'' and TS 3/4.3.3.9 from
the TS to the UFSAR. Movable Incore Detectors and Radioactive
Gaseous Effluent Oxygen Monitoring Instrumentation are not
initiators to any accident previously evaluated. Consequently, the
probability of an accident previously evaluated is not significantly
increased. Movable Incore Detectors and Radioactive Gaseous Effluent
Oxygen Monitoring Instrumentation are not accident mitigating
structures, systems, or components. No impact on the plant response
to accidents will be created. Thus the consequences of accidents
previously analyzed are unchanged between the existing TS
requirements and the proposed changes.
The proposed revision to TS SR [Surveillance Requirement]
4.11.2.5 is an administrative change to a reference necessitated by
the proposed relocation of TS Table 3.3-13 from the TS to the UFSAR.
The proposed revision to the TS Index, page renumbering, and minor
format changes to improve consistency are also administrative
changes necessitated by the proposed relocation of TS 3/4.3.3.2 and
TS 3/4.3.3.9 from the TS to the UFSAR.
Therefore, the proposed changes do not involve a significant
increase in the probability or radiological consequences of an
accident previously evaluated.
2. Do the proposed change[s] create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated in
the UFSAR. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. Specifically, no new hardware is being added to the plant
as part of the proposed changes, no existing equipment is being
modified, and no significant changes in operations are being
introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed change[s] involve a significant reduction in
a margin of safety?
Response: No.
The proposed changes will not alter any assumptions, initial
conditions, or results of any accident analyses. The Movable Incore
Detectors and oxygen monitoring instrumentation will continue to
perform as before. The proposed changes relocate TS 3/4.3.3.2 and TS
3/4.3.3.9 from the TS to the UFSAR consistent with the guidance in
NRC Generic Letter 95-10 and 10 CFR 50.36, and make conforming
administrative changes to the TS Index, page renumbering, and minor
format changes to improve consistency and to TS SR 4.11.2.5 to
reflect the relocation of TS 3/4.3.3.9 from the TS to the Salem
UFSAR.
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.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendments request: September 26, 2006.
Description of amendments request: The amendments would revise
Technical Specification 6.9.1.9 to remove the revision number and date
for the topical reports that contain the analytical methods used in the
Core Operating Limits Report (COLR). The effect of this change is to
allow the licensee to use current topical reports, as long as they have
been approved by the NRC. The amendments would also add an NCR-approved
topical report to the Salem Nuclear Generating Station, Unit No. 2,
COLR methods.
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.
The proposed changes affect the administrative controls section
of Technical Specifications (TS) that govern the analytical methods
used to determine core operating limits. Removal of revision levels
and dates from NRC-approved methods listed in TS is
[[Page 65144]]
an administrative change that has no impact on the probability or
consequences of an accident. TS 6.9.1.9.b will still require these
methods to be reviewed and approved by [the] NRC. The proposed
change does not affect the required TS actions to be taken in the
event that any core operating limits are exceeded.
The proposed use of WCAP-10054-P-A, Addendum 2 for the Salem
Unit 2 Small Break Loss of Coolant Accident (SBLOCA) analysis is
consistent with the limitations and conditions of NRC approval. The
parameters assumed in the analysis are within the design limits of
the plant equipment. Therefore, there will be no increase in the
probability of a loss of coolant accident. The consequences of a
LOCA are not being increased, since it is shown that the Emergency
Core Cooling System (ECCS) is designed so that its calculated
cooling performance conforms to the criteria contained in 10 CFR
50.46, Paragraph b. No other accident is potentially affected by
this change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new modes of plant operation are being introduced. The
parameters assumed in the analysis are within the design limits of
the plant equipment. TS will continue to require operation within
the core operating limits determined using NRC-approved analytical
methods and the proposed change does not affect any actions required
in the event the core operating limits are exceeded.
Therefore, the proposed change does not involve an increase in
the probability or consequences of an accident previously evaluated.
3. Do the proposed change[s] involve a significant reduction in
a margin of safety?
Response: No.
The proposed changes do not have any impact on plant equipment
or safety analysis acceptance criteria. Core operating limits will
continue to be determined using NRC-approved analytical methods. The
ECCS acceptance criteria of 10 CFR 50.46 will continue to be met
following the proposed use of WCAP-10054-P-A, Addendum 2 for the
Salem Unit 2 SBLOCA analysis[.]
Therefore, the proposed change[s] do[es] 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.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of amendment request: April 6, 2006.
Description of amendment request: The amendment would change the
Technical Specifications to reduce the maximum allowable reactor power
when two main steam safety valves (MSSVs) are inoperable.
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[es] the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change reduces the power level at which Salem Unit
2 may be operated with a maximum of two inoperable MSSVs in any
steam generator. This change is consistent with analyses of the
limiting transients for secondary system pressure (loss of load/
turbine trip and rod withdrawal at power), performed to demonstrate
the acceptance criterion of 110% of design pressure will continue to
be met following steam generator replacement. The proposed change
does not involve any changes to the MSSV actuation setpoints; they
remain well above the Main Steam System operating pressures. The
proposed change does not challenge the relief capacity of the MSSVs.
Therefore, the probability of an event associated with mis-operation
of the MSSVs (e.g., inadvertent depressurization of the Main Steam
System) is not impacted by the proposed change. The proposed
reduction in allowable power level establishes initial conditions
consistent with the safety analyses, and does not affect the
probability of any previously evaluated accident.
The proposed change is necessitated by analyses of limiting
secondary system pressure transients, whose acceptance criteria
continue to be met provided that plant operation is restricted to
58% RTP [rated thermal power] with a maximum of two inoperable MSSVs
in any steam generator. There is no impact on any radiological
consequences of an accident associated with the proposed reduction
in maximum power level.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Do[es] the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Reducing the allowable power level per the proposed change does
not introduce any new accident scenarios or malfunctions. The
proposed change establishes an operating restriction consistent with
current safety analysis methodology. It represents a change to an
input assumption used in analyses of limiting secondary
pressurization transients to ensure plant operation does not
challenge any design limits.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do[es] the proposed change involve a significant reduction in
a margin of safety?
Response: No.
Acceptable margins of safety are inherent in the safety analysis
acceptance criteria, including the limit on secondary system
pressure to 110% of design pressure during a loss of load/turbine
trip (LOL/TT) or rod withdrawal at power (RWAP) transient. The
purpose of the proposed change is to limit operation with a maximum
of two inoperable MSSVs for any steam generator, such that the
acceptance criterion for secondary pressure continues to be met. The
proposed change does not modify any acceptance criteria, nor would
it cause any design limit to be exceeded.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: September 29, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.7.8, ``Service Water (SW) System,'' to
change the limiting conditions for operation (LCOs), Actions,
Completion Times, and Surveillance Requirements (SRs). Specifically,
the proposed amendment would change the LCO to require a specific
number of SW pumps to be operable rather than the current SW train
operability. The LCO Actions, Completion Times, and SRs would also be
revised based on pump operability.
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:
[[Page 65145]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The safety related function of the Service Water (SW) System is
to provide cooling for safety related equipment, mitigate the
containment response effects of a Main Steam Line Break (MSLB) and
design basis Loss of Coolant Accident (LOCA), and provide long term
containment and core cooling in the event of a LOCA. The operation
of the SW system, including the number of pumps operating or
available, has no affect on the probability of these accidents.
The probability of a loss of SW event is not increased. The
proposed TS provides for more restrictive actions for pump
inoperability than the existing TS, thereby reducing the probability
of this event.
The consequences of a[n] MSLB or LOCA or other design basis
accidents are not increased beyond that assumed in the accident
analysis. Two service water pumps are sufficient for all accident
mitigation functions. The change provides for adequate service water
supply (2 pumps) for both normal and accident conditions. The
availability of associated power supplies is also considered. For a
reduction in the total number of available pumps, appropriate LCO
action statements ensure that the pumps are returned to service
within a time limit commensurate with an acceptable level of plant
safety and risk, or the plant is placed in a safe mode.
The loss of SW has been previously evaluated and measures
implemented to mitigate the event. Since a loss of SW assumes no SW
pumps are operating, the proposed amendment has no affect on
consequences of this event.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The only accidents directly initiated from this system are the
loss of SW or flooding concerns. Both of these accidents have been
previously evaluated with acceptable results. Therefore, this change
does not create the possibility of a new or different [kind] of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change will ensure that sufficient SW pumps are available
for accident mitigation at any one time while still providing the
appropriate operational flexibility. A risk determination
demonstrates that any increase in risk associated with this change
is within the established regulatory guidelines. The technical
analysis shows that appropriate action statements exist to ensure
adequate SW is available for accident mitigation, considering
emergency power supply availability. Therefore, this proposed change
does not involve a significant reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Richard J. Laufer.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: October 12, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.3.3, ``Capacity,'' to change the
limit on the number of fuel assemblies in the spent fuel pool. The
proposed amendment would also revise TS 3.7.13, ``Spent Fuel Pool
Storage,'' to remove the references to Type 4 spent fuel pool storage
racks, which are not currently installed.
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 reduces the total number of fuel assemblies
that can be stored in the current spent fuel pool storage locations
and reduces the number of available locations. This will limit the
potential inventory of spent fuel in the pool. The probability of an
accident has not changed since the number of stored fuel assemblies
is not a precursor for a spent fuel handling accident. A comparison
of the criticality analysis of fuel assemblies to be used in
subsequent Extended Power Uprate core reloads to the current
criticality analysis has been performed. The design parameter
assumptions used in the licensing basis criticality analyses are
bounding.
There are no new components or new functions associated with the
spent fuel cooling system so the probability of an accident has not
changed. The effect of a single failure on the spent fuel pool
system's capability to provide for heat removal from the fuel pool
has been analyzed. The analysis concluded that the system remains
within the parameters previously evaluated. The implementation of
the Extended Power Uprate does not affect the capability of the
system to perform its function.
The Extended Power Uprate conditions do not add any new or
previously unevaluated materials to the spent fuel pool storage
system and do not include any reductions in the boron concentration
requirements so the probability of an accident has not changed. The
total soluble boron concentration required to maintain the spent
fuel pool in a subcritical condition with the transition to the new
fuel has not changed. The conclusions in the Ginna UFSAR [Updated
Final Safety Analysis Report], assuming the most limiting accident,
remain valid.
Therefore, the consequences of a fuel handling accident, a loss
of spent fuel cooling, and a boron reduction concentration event
previously evaluated have not increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not alter the function of the spent fuel
pool or any related equipment, nor cause it to operate differently
than it was designed to operate. All equipment required to mitigate
the consequences of an accident would continue to operate as before.
The proposed changes reduce the maximum number of fuel assemblies
that can be stored in the spent fuel pool and the number of storage
locations. Therefore, this change does not create the possibility of
a new or different [kind] of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes reduce the maximum number of fuel
assemblies that can be stored in the spent fuel pool and the number
of storage locations. The changes are in accordance with conclusions
supporting Extended Power Uprate and have been determined to be
acceptable. The design parameter assumptions used in the licensing
basis criticality analysis bound those of the new fuel assemblies.
Although the individual heat load per assembly has increased due to
the changed fuel design, the maximum spent fuel pool heat load has
decreased due to the reduction in the number of fuel assemblies that
will be stored based on future plans to use dry cask storage.
Therefore, this proposed change does not reduce 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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Richard J. Laufer.
[[Page 65146]]
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: October 3, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specifications surveillance requirements related
to inspection of the containment sump trash racks and screens, inside
recirculation spray (RS) pump wells, and outside RS and low head safety
injection pump suction inlets resulting from Nuclear Regulatory
Commission's (NRC's) Generic Safety Issue (GSI) 191 and Generic Letter
(GL) 2004-02.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not impact the condition or performance
of any plant structure, system or component. Furthermore, the
proposed change does not affect the initiators of any previously
analyzed event or the assumed mitigation of accident or transient
events since the plant will be operated in the same manner and
within the same operating limits that are currently in place. The
proposed TS change is administrative in nature given that inspection
of containment sump components for debris accumulation and
structural integrity will continue to be performed. The revised TS
surveillance wording is being implemented as a clarification to
facilitate inspection of the containment sump in its current
configuration, as well as after containment sump modifications have
been implemented in response to GSI-191 and GL 2004-002. As a
result, the proposed change to the Surry TS does not involve any
increase in the probability or the consequences of any accident or
malfunction of equipment important to safety previously evaluated
since neither accident probabilities nor consequences are being
affected by this proposed change.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change is administrative in nature and, as such,
does not involve any changes in station operation or physical
modifications to the plant. In addition, no changes are being made
in the methods used to respond to plant transients that have been
previously analyzed. No changes are being made to plant parameters
within which the plant is normally operated or in the setpoints,
that initiate protective or mitigative actions, since the plant will
be operated in the same manner and within the same operating limits
that are currently in place. Since plant operation will not be
affected by this change, no new failure modes are being introduced.
Therefore, the proposed change to the Surry TS does not create the
possibility of a new or different kind of accident or malfunction of
equipment important to safety from any previously evaluated.
3. Does the change involve a significant reduction in the margin
of safety?
The proposed change is administrative in nature given that
inspection of the containment sump components for debris
accumulation and structural integrity will continue to be performed
on an established frequency. The more general nature of the TS
surveillance wording is being implemented as a clarification to
facilitate inspection of the containment sump in its current
configuration, as well as after containment sump modifications have
been implemented in response to GSI-191 and GL 2004-002. The
proposed change does not impact station operation or any plant
structure, system or component that is relied upon for accident
mitigation. Furthermore, the margin of safety assumed in the plant
safety analysis is not affected in any way by the proposed change
since the plant will be operated in the same manner and within the
same operating limits and setpoints that are currently in place.
Therefore, the proposed change to the Surry Technical Specifications
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
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 51.12(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
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].
Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: November 30, 2005.
Brief description of amendment: The amendment revises the
surveillance requirements (SR) for the emergency diesel generator
automatic trips bypass of SR 3.8.1.11 from 18 months to 24 months.
Date of issuance: October, 4, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No. 208.
Renewed Facility Operating License No. DPR-23. Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10072).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 2006.
[[Page 65147]]
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
Duke Power Company LLC, et al., Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: July 27, 2005, as supplemented
May 4, 2006, and August 8, 2006.
Brief description of amendments: The amendments revise the Catawba
and McGuire Technical Specification 3.4.15, ``RCS Leakage Detection
Instrumentation.''
Date of issuance: September 30, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 234/230 and 235/217.
Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and
NPF-17: Amendments revised the licenses and the technical
specifications.
Date of initial notice in Federal Register: August 30, 2006 (71 FR
51644).
The supplement dated August 8, 2006, provided clarifying
information that did not expand the scope of the July 27, 2005,
application as modified May 4, 2006.
The Commission's related evaluation, Final No Significant Hazards
Finding, and State consultation of the amendments are contained in a
Safety Evaluation dated September 30, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. STN 50-457, Braidwood
Station, Unit No. 2, Will County, Illinois
Date of application for amendment: November 18, 2005, as
supplemented by letters dated August 18 and September 28, 2006.
Brief description of amendment: The amendment revised TS 5.5.9,
``Steam Generator (SG) Tube Surveillance Program,'' regarding the
required SG inspection scope for Braidwood Station, Unit No. 2, during
refueling outage 12 and the subsequent operating cycle.
Date of issuance: October 24, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 141.
Facility Operating License No. NPF-77: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: (71 FR 29676; May 23,
2006).
The August 18 and September 28, 2006, supplements contained
clarifying information and did not change the NRC staff's initial
proposed finding of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 24, 2006.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 23, 2006.
Description of amendment request: The amendment deleted License
Condition 2.G, ``Reporting to the Commission,'' as described in the
Notice of Availability published in the Federal Register on April 25,
2006 (71 FR 23955). The change was requested as part of the
consolidated line item improvement process and consistent with the
model safety evaluation published in the Federal Register on November
4, 2005 (70 FR 67202).
Date of issuance: October 17, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 113.
Facility Operating License No. NPF-86: The amendment revised
Facility Operating License No. NPF-86 and the Technical Specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23955).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 17, 2006.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 16, 2006.
Brief description of amendment: The amendment revised the Technical
Specification 3.10.1, ``Inservice Leak and Hydrostatic Testing
Operation,'' to extend the scope to include provisions for temperature
increases above 212 [deg]F as a consequence of inservice leak or
hydrostatic testing, and as a consequence of control rod scram time
testing initiated in conjunction with the inservice leak test or
hydrostatic test, when initial test conditions are below 212 [deg]F.
Date of issuance: October 23, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 225.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43535)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 23, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: December 7, 2005, as
supplemented by letters dated July 20 and September 5, 2006.
Brief description of amendments: These amendments revised the
Technical Specifications to delete Surveillance Requirement (SR)
4.9.2.b, which requires performance of a channel functional test (CFT)
of each source range neutron flux monitor within 8 hours prior to the
initial start of core alterations. An associated administrative change
would renumber current SR 4.9.2.c as SR 4.9.2.b. The amendments would
also eliminate the restriction in SRs 4.10.3.2 and 4.10.4.2 that the
CFTs of the intermediate and power range monitors be performed within
12 hours prior to initiating physics tests.
Date of issuance: October 13, 2006.
Effective date: As of the date of issuance, to be implemented in 60
days.
Amendment Nos.: 275, 257.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: August 2, 2006 (71 FR
43819). The supplements 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 October 13, 2006.
No significant hazards consideration comments received: No.
[[Page 65148]]
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: November 15, 2005, as
supplemented May 31, August 31, and September 29, 2006.
Brief description of amendment: The amendment revises the Virgil C.
Summer Nuclear Station Technical Specifications (TS) 3/4.3 for the
reactor trip instrumentation and the engineered safety feature
actuation system instrumentation to implement the allowed outage time
and bypass test time changes approved in WCAP-14333-P-A, Revision 1,
``Probabilistic Risk Analysis of the RPS and ESFAS Test Times and
Completion Times,'' and makes several additional changes to TS outside
of the scope of WCAP-14333.
Date of issuance: October 24, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No. 177.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75496).
The supplemental letters provided clarifying information that was
within the scope of the initial notice and did not change the initial
proposed no significant hazards consideration. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
October 24, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: November 10, 2003 (TS-430), as
supplemented by letter dated November 8, 2004.
Brief description of amendment: The amendment incorporates the
necessary Technical Specification (TS) changes for the planned
replacement of the power range monitoring portion of the existing
Neutron Monitoring System with a digital upgrade. These changes expand
the current allowable operating domain to the Maximum Extended Load
Line Limit region of the power/flow chart.
Date of issuance: September 27, 2006.
Effective date: Date of issuance, to be implemented within 30 days.
Amendment No.: 262.
Facility Operating License No. DPR-33: Amendment revised the TSs.
Date of initial notice in Federal Register: February 3, 2004 (69 FR
5208). The November 8, 2004, supplement, contained clarifying
information and did not change the NRC staff's initial proposed finding
of no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 27, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 6, 2006.
Brief description of amendments: The amendments modify Technical
Specification (TS) requirements for inoperable snubbers by adding
Limiting Condition for Operation 3.0.7. This operating license
improvement was made available by the Nuclear Regulatory Commission
(NRC) on May 4, 2005 (70 FR 23252) as part of the consolidated line
item improvement process and is consistent with NRC approved Technical
Specification Task Force (TSTF) standard TS change TSTF-372, Revision
4.
Date of issuance: October 4, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos. 312/301.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15487).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 4, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 30th day of October 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-18595 Filed 11-6-06; 8:45 am]
BILLING CODE 7590-01-P