[Federal Register Volume 71, Number 39 (Tuesday, February 28, 2006)]
[Notices]
[Pages 10071-10084]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-1737]
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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 February 3, 2006, to February 15, 2006. The
last biweekly notice was published on February 14, 2006 (71 FR 7804).
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, Rules and
Directives 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 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.
[[Page 10072]]
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:
Rulemaking 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 email 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].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: November 30, 2005.
Description of amendment request: The proposed amendment would
revise the frequency of the diesel generator automatic trips bypass
surveillance requirement (SR) 3.8.1.11 from 18 months to 24 months.
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?
No. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change decreases the frequency of SR
3.8.1.11, verification of the DG [diesel generator] automatic trips
bypass, from 18 months to 24 months. The DG automatic trips bypass
circuitry is required for DG operability and reliability during
emergency operation of the DG. The proposed test frequency will
continue to assure that the DG will perform as required. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated,
because the factors that are used to determine the probability and
consequences of accidents are not being affected.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated. There
are no new or different accident initiators or sequences being
created by the proposed Technical Specifications change. The
required surveillance performed at the proposed frequency will
continue to provide assurance that the trips bypass function is
operable and is properly supporting operation of the associated DG.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
No. The proposed change does not involve a significant reduction
in the margin of
[[Page 10073]]
safety. The proposed change will continue to ensure that the DG
trips bypass function operates as designed. The functionality and
operability of emergency power system is not being changed.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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: Michael L. Marshall, Jr.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: January 4, 2006.
Description of amendment request: The proposed amendment would
change the Millstone Power Station, Unit No. 2 Technical Specification
(TS) 3/4.3.3.8, ``Instrumentation, Accident Monitoring,'' to modify the
description of the pressurizer power operated relief valves (PORVs) and
pressurizer safety valves position indicators.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment removes the wording ``Acoustic Monitor,''
which provides specific details related to system design, from items
4 and 6 of TS 3/4.3.3.8, Tables 3.3-11 and 4.3-7. The PORVs and
Pressurizer Safety Valves position indicators (and the associated
``Acoustic Monitor'') provide only indications of valve position.
They do not constitute a design feature that is an initial condition
for a design basis accident or transient analysis. Furthermore, they
do not affect the function of the system, equipment in the system or
actuate to mitigate a design basis accident or transient. Therefore,
the proposed changes do not increase the probability or consequences
of an accident previously evaluated.
Additionally, the TS retains the requirement for the total and
minimum channels required to be OPERABLE and to verify channel
OPERABILITY at the designated frequencies. The PORVs and Pressurizer
Safety Valves are equipped with positive position indication that
meets the requirements of RG [Regulatory Guide] 1.97.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not impact the capability of existing
equipment to perform its intended functions. No system setpoints are
being modified and no changes are being made to the method in which
plant operations are conducted. No new failure modes that would
impact accident analyses are introduced by the proposed changes. The
proposed amendment does not introduce 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 proposed amendment removes the wording ``Acoustic Monitor''
from items 4 and 6 of TS 3/4.3.3.8, Table[s] 3.3-11 and 4.3-7. The
proposed changes do not affect any of the assumptions used in the
accident analysis, nor does it affect any operability requirements
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: Darrell J. Roberts.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 30, 2005.
Description of amendment request: The proposed amendment
establishes a combined leakage rate limit for the sum of the four Main
Steam line leakage rates that is equal to four times the current
individual Main Steam Isolation Valve (MSIV) leakage rate limit.
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 amendment does not involve a change to structures,
systems, or components that would affect the probability of an
accident previously evaluated in the Cooper Nuclear Station (CNS)
Updated Safety Analysis Report (USAR). The proposed amendment
results in no change in the radiological consequences of the design
basis Loss-of-Coolant Accident (LOCA) as currently analyzed for CNS.
That analysis was calculated for a combined Main Steam Isolation
Valve (MSIV) leakage for determining acceptance to the regulatory
limits for the offsite and Control Room radiation doses, as
contained in 10 CFR 100 [Part 100 of Title 10 of the Code of Federal
Regulations] and 10 CFR 50[,] Appendix A, General Design Criterion
(GDC) 19. The aggregate Main Steam line leakage rate limit has no
adverse effect on the environmental qualification of equipment
important to safety, as provided for in 10 CFR 50.49.
Based on the above conclusions, this proposed amendment 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.
The proposed change does not modify the MSIVs or any other plant
system or structure associated with this amendment and therefore,
will not affect their capability to perform their design function.
The combined total Main Steam line leakage rate is included in the
current radiological analyses for the assessment of radiation
exposure following an accident. This License Amendment Request
revises the allowable leakage rate from a per valve limit to a total
combined leakage rate limit for all four Main Steam lines but does
not change the cumulative limit.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The leakage rate limit specified for the MSIVs is used to
quantify the maximum amount of Secondary Containment bypass leakage
assumed in the LOCA radiological analysis. Results of the analysis
are evaluated against the dose limits contained in 10 CFR 50[,]
Appendix A[,] GDC 19 and 10 CFR 100. The margin of safety in this
context is considered to be the difference between the calculated
dose exposures and the limits provided by GDC 19 and 10 CFR 100.
Therefore, since the proposed combined Main Steam line leakage
rate limit is unchanged from the assumed maximum leakage rate for
MSIVs, for the purpose of calculating [a] potential radiation dose,
the margin of safety is not affected because the postulated
radiation doses remain the same.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 10074]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: David Terao.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 30, 2006.
Description of amendment request: The proposed change allows a
delay time for entering a supported system Technical Specification (TS)
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed consistent with the program in place for
complying with the requirements of 10 CFR 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.8 is added to the TS to provide this
allowance and define the requirements and limitations for its use.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
November 24, 2004 (69 FR 68412), on possible amendments concerning
TSTF-372, including a model safety evaluation and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 4, 2005 (70 FR 23252). The
licensee affirmed the applicability of the following NSHC determination
in its application dated January 30, 2006.
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 change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on
allowance provided by proposed LCO 3.0.8 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by this change. The addition of a
requirement to assess and manage the risk introduced by this change
will further minimize possible concerns. Therefore, this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in Regulatory Guide 1.177. A bounding risk assessment
was performed to justify the proposed TS changes. The proposed LCO
3.0.8 defines limitations on the use of the provision and includes a
requirement for the licensee to assess and manage the risk
associated with operation with an inoperable snubber. 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. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: David Terao.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: November 11, 2005.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.6.5, ``Containment Spray and
Cooling Systems''; an existing Condition, two Surveillance
Requirements, and add a new Condition which will allow continued plant
operation with TS limitations when two Containment Cooling System fan
coil units (FCUs), one in each train, 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 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 to allow plant operation to continue for a limited
time period under Technical Specification controls with two fan coil
units, one fan coil unit from each containment cooling train,
providing the required cooling function. Analyses demonstrate that
any two fan coil units, whether they are in the same train or from
opposite trains, are sufficient to supply the required containment
cooling following a design basis accident when the plant in the
proper configuration as required by the proposed Technical
Specifications.
The containment cooling system is required for accident
mitigation and is not an accident initiator, thus revising the
equipment required to provide the safety function does not involve a
significant increase in the probability of an accident previously
evaluated.
Since the proposed change continues to provide the post-accident
containment cooling function under Technical Specification controls,
this change does not involve an increase in the consequences of an
accident. Thus this 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.
This license amendment proposes to revise the Technical
Specifications to allow plant operation to continue for a limited
time period under Technical Specification controls with two fan coil
units, one fan coil unit from each containment cooling train,
providing the required cooling function. Analyses demonstrate that
any two fan coil units, whether they are in the same train or
[[Page 10075]]
from opposite trains, are sufficient to supply the required
containment cooling following a design basis accident when the plant
in the proper configuration as required by the proposed Technical
Specifications.
The proposed licensing basis changes do not involve a change in
the function or use of the containment cooling system. It does
assure that the containment cooling function is provided during
plant operations for post-accident mitigation. There are no new
failure modes or mechanisms created through allowing different
combinations of fan coil units to provide the cooling function as
proposed by this Technical Specification change. There are no new
accident precursors generated by providing the required cooling
function with an operable fan coil unit from each train.
Therefore, 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.
This license amendment proposes to revise the Technical
Specifications to allow plant operation to continue for a limited
time period under Technical Specification controls with two fan coil
units, one fan coil unit from each containment cooling train,
providing the required cooling function. Analyses demonstrate that
any two fan coil units, whether they are in the same train or from
opposite trains, are sufficient to supply the required containment
cooling following a design basis accident when the plant in the
proper configuration as required by the proposed Technical
Specifications.
Current plant Technical Specifications allow plant operation to
continue for 7 days with the containment cooling function provided
by the two operable fan coil units of a single operable containment
cooling train. This is acceptable because engineering analyses
demonstrate that the two fan coil units of a single train can
provide the required post-accident containment cooling.
Likewise, engineering analyses demonstrate that any two fan coil
units from opposite containment cooling trains can also provide the
required post-accident containment cooling if the cooling water flow
to the other fan coil unit in each train is isolated. This license
amendment request proposes Technical Specifications which will allow
plant operation to continue for 7 days with the containment cooling
function provided by two fan coils from opposite trains provided the
cooling water flow to the other fan coil unit in each train is
isolated. Thus, from a cooling capacity perspective, this proposed
Technical Specification change does not involve a reduction in a
margin of safety.
When inoperable plant systems are under Technical Specification
controls that limit the time for inoperability, a single failure in
addition to the inoperable equipment is not postulated. Therefore,
whether two inoperable fan coil units are in the same train or
opposite trains does not change the availability of the two
remaining operable fan coil units. Thus from a Technical
Specification perspective, this proposed Technical Specification
change does not involve a reduction in a margin of safety.
Therefore, based on the considerations given above, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Timothy J. Kobetz.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 19, 2005.
Description of amendment request: The proposed change will revise
Fort Calhoun Station, (FCS) Technical Specification 2.4, ``Containment
Cooling,'' (and associated Bases) to reduce the required number of
operable Containment Spray (CS) pumps from three to two in order to
enhance net positive suction head (NPSH) margins. This change will be
accomplished by disabling the containment spray actuation signal (CSAS)
automatic start feature of CS pump SI-3C. This change will reduce the
head loss across the containment sump strainers during the
recirculation phase of a design-basis accident (DBA) by reducing flow
rates, and will improve NPSH available (NPSHA).
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 Containment Spray (CS) system is not an initiator of any
accident previously evaluated at the Fort Calhoun Station (FCS); the
CS system is an accident mitigation system. The CS system's
licensing basis functions are to limit the containment pressure rise
and reduce the leakage of airborne radioactivity from the
containment by providing a means for cooling the containment
following a loss-of-coolant accident (LOCA) or main steam line break
(MSLB) inside containment. The proposed change disables the CSAS
automatic start feature of one of the three CS pumps.
The only FCS safety analysis that currently assumes three CS
pumps operating to mitigate an accident is the Containment Pressure
Analysis for a[n] MSLB inside containment. Even though this analysis
assumes operation of all three CS pumps, it also shows that peak
containment pressure occurs prior to the CS system starting,
therefore, the CS system does not mitigate the peak pressure for
a[n] MSLB. The reviews evaluated both existing AORs [analyses of
record] and those analyses developed for the Steam Generator
Replacement (RSG) project. The analysis developed for the RSG
project that evaluates the Containment Pressure Analysis for MSLB
inside containment was reviewed for the impact of reducing the
number of operating CS pumps from three to two. This review
determined that the RSG MSLB analysis will be acceptable and will
continue to be bounded by the analysis currently documented in USAR.
AOR peak pressure is unaffected by implementation of this proposed
change. Therefore, the combination of the RSG project and this
containment spray modification will not result in an increase in the
currently documented peak containment pressure for an MSLB.
Therefore, the evaluation for the MSLB event has determined that the
containment pressure response is acceptable with less than three CS
pumps operating.
The LOCA analysis source term is based on operation of minimum
safeguards due to a worst-case single failure. The minimum
safeguards configuration is unchanged by this modification.
Following implementation of the proposed change at least one CS pump
will be available to mitigate a LOCA as currently assumed in the
analysis, therefore, the proposed change will have no adverse effect
on the radiological consequences following a LOCA. The analyses that
establish the radiological consequences for the site are based on a
Large Break LOCA with a single CS pump in operation, therefore,
single CS pump operation during a[n] MSLB inside containment is
bounded by the LOCA analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change will reduce the number of operable CS pumps
from three to two; however, previous accident analyses will remain
valid. No credible new failure mechanisms, malfunctions, or accident
initiators not considered in the design and licensing basis have
been created and none of the initial condition assumptions of any
accident evaluated in the safety analysis are impacted.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
[[Page 10076]]
Response: No.
The containment building and associated penetrations are
designed to withstand an internal pressure of 60 psig at 305 [deg]F,
including all thermal loads resulting from the temperature
associated with this pressure, with a leakage rate of 0.1 percent by
weight or less of the contained volume per 24 hours. The CS System
and the Containment Fan Coolers are credited for maintaining
containment pressure and temperatures within design limitations, and
assure that the release of fission products to the environment
following a design[-]basis accident will not exceed regulatory
guidelines. The FCS licensing basis credits only one of the three CS
pumps to limit the containment pressure to below the design value
for a LOCA. Currently, the FCS licensing basis credits three CS
pumps for a[n] MSLB, however, the CS system is not credited for
limiting peak containment pressure for a[n] MSLB.
The EEQ [electrical equipment qualification] profile developed
for the current plant configuration bounds those associated with the
upcoming RSG modification. Both the proposed CS system changes and
the RSG projects are scheduled for the same refueling outage. The
thermal lag analysis of equipment performed using the current plant
configuration demonstrated a large margin between the equipment
evaluated during the accident versus the conditions under which it
was tested. The RSG modification will further increase this margin.
As part of the RSG effort the EEQ analysis will be revised to
address RSG issues and will include the changes to containment
spray. When the margins associated with the current analysis as well
as increases in margin when the new analysis is implemented it is
expected that the changes to the containment spray system will not
produce an adverse result. All equipment will remain qualified to
operate in the accident environment.
Additionally, the CFCs [containment fan coolers] operate
independently of the CS system to remove heat from the containment
atmosphere. The CFCs consist of two redundant trains; each train
with one air cooling and filtering unit and one air cooling unit,
for a total of four cooling units. Operation of the CFCs is credited
in the MSLB containment pressure analysis. The CFCs are not impacted
by this proposed change. During the MSLB containment spray takes
place after the peak containment pressure occurs. Therefore, the
licensing basis capabilities of the Containment Cooling System,
which consists of the CS and CFCs, is not adversely affected by the
proposed change; the ability to maintain containment peak pressure
and temperature and long[-]term containment pressure and temperature
will be maintained.
Particulate fission products that are released into the
containment following a DBA are removed by the CS system for those
events that result in CS actuation. The water spray strips
radioactive particles from the atmosphere where they fall to the
floor and are washed into the containment sump. The radiological
consequences analysis credits CS system operation for removal of
particulates from the containment atmosphere during a LOCA. The LOCA
analysis source term is based on operation of minimum safeguards due
to a worst-case single failure, and a presumption of core damage.
Minimum safeguards corresponds to one CS pump and one CS header
operation and take into account pump degradation, and instrument
uncertainties. The analyses that establish the radiological
consequences for the site are not impacted by the proposed
modification. These analyses are based on a Large Break LOCA with a
single CS pump in operation. Therefore, single CS pump operation
bounds the plant configuration following the proposed modification.
The Large Break LOCA assumes that there will be three CS pumps
operating when evaluating the effects of containment pressure on
ECCS [emergency core cooling system] performance. The analysis
assumes three CS pumps, which minimizes containment pressure, to
conservatively evaluate ECCS performance in response to a LOCA. The
use of two CS pumps versus three improves ECCS performance and thus
increases margin to 10 CFR 50.46 limits on peak clad temperature.
In summary, following implementation of the proposed change:
Peak containment pressure for analyzed DBAs will not be
increased;
The assumptions used in the environmental qualification
of equipment exposed to the containment atmosphere following a DBA
remaining bounding; and
The radiological consequences for the bounding DBA
remains unchanged.
The currently calculated peak clad temperature
following a LOCA remains bounded by existing analysis.
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: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Branch Chief: David Terao.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California
Date of amendment requests: January 13, 2006.
Description of amendment requests: The proposed amendment would
revise Technical Specification 5.6.5, ``Core Operating Limits Report
(COLR),'' by adding WCAP-16009-P-A, ``Realistic Large-Break LOCA [Loss-
of-Coolant Accident] Evaluation Methodology Using the Automated
Statistical Treatment of Uncertainty Method (ASTRUM),'' dated January
2005, as an approved analytical method for determining core operating
limits for Unit 2.
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 to allow the use of the best estimate loss-
of-coolant accident (LOCA) analysis methodology using the automated
statistical treatment of uncertainty methodology (ASTRUM) does not
involve a physical alteration of any plant equipment or change
operating practice at Unit 2 of Diablo Canyon Power Plant (DCPP).
Therefore, there will be no increase in the probability of a LOCA.
The consequences of a LOCA are not being increased.
The plant conditions assumed in the analysis are bounded by the
design conditions for all equipment in Unit 2. That is, it is shown
that the emergency core cooling system is designed so that its
calculated cooling performance conforms to the criteria contained in
10 CFR [Title 10 of the Code of Federal Regulations, Section] 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. Does the proposed change create the possibility of a new or
different accident from any accident previously evaluated?
Response: No.
The proposed change would not result in any physical alteration
to any Unit 2 system, and there would not be a change in the method
by which any safety [-]related system performs its function.
Analyses of transient events have confirmed that no transient event
results in a new sequence of events that could lead to a new
accident scenario. The parameters assumed in the analysis are within
the design limits of existing plant equipment.
In addition, employing the ASTRUM methodology does not create
any new failure modes that could lead to a different kind of
accident. The design of all systems remains unchanged and no changes
are being made to any reactor protection system or engineered
safeguard features actuation setpoints.
Based on this review, it is concluded that no new accident
scenarios, failure mechanisms or limiting single failures are
introduced as a result of the proposed changes.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
[[Page 10077]]
It has been shown that the analytic technique used in the
analysis realistically describes the expected behavior of the DCPP
Unit 2 reactor system during a postulated LOCA. Uncertainties have
been accounted for as required by 10 CFR 50.46. A sufficient number
of LOCAs with different break sizes, different locations, and other
variations in properties have been analyzed to provide assurance
that the most severe postulated LOCAs were analyzed. The analysis
has demonstrated that all acceptance criteria contained in 10 CFR
50.46[,] paragraph b continue to be satisfied.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California
Date of amendment request: January 19, 2006.
Description of amendment request: The licensee has proposed to
revise the Technical Specifications (TS) to correct an editorial error
in TS 3.1.2, ``Spent Fuel Pool Load Restrictions,'' and to change TS
5.2.2, ``Facility Staff,'' to allow the Unit 3 control room to be
temporarily unmanned during emergency conditions that require personnel
to evacuate buildings for their safety.
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 editorial change has no impact on probability or
consequences of accidents. The following discussion applies to the
proposed change related to control room evacuation.
Allowing plant personnel to not continuously man the control
room has no impact on the probability of an accident from occurring,
especially acts of nature such as earthquakes and tsunamis.
The HBPP DSAR, Appendix A, and NRC SER, Section 10, dated April
29, 1987, evaluate various accidents at HBPP. Because all fuel has
been removed from the reactor vessel and stored in the spent fuel
pool, the majority of accidents analyzed pertain to events that
could only affect spent fuel or the spent fuel pool. All accidents
affecting spent fuel or the spent fuel pool do not require operator
action to protect the public health and safety or to maintain
offsite radiological doses well within regulatory limits. In
addition, NRC SER, Section 10.7, ``Impact of Tsunami Flooding,''
analyzes the impact of tsunami flooding. That analysis identifies a
likely impact of the tsunami to be a release of the radwaste tank
radionuclide contents to the bay and some damage to the reactor
building. For both situations, no operator action is required to
maintain offsite radiological doses well within regulatory limits.
Allowing the control room to be temporarily unmanned under
emergency conditions does not create problems that could increase
the consequences of an accident. The primary function of manning the
control room is for an operator to observe and acknowledge alarms.
Recovery actions to respond to damage to spent fuel, the spent fuel
pool, or radwaste tanks are taken by personnel outside the control
room. No recovery actions are required to be taken by the control
room operator to respond to damage to spent fuel, the spent fuel
pool, or radwaste tanks.
Evacuating occupied buildings, including the control room,
during a tsunami, allows the control room operator to return to the
control room after the tsunami and assess damage by observing
indicators and alarms. Upon returning to the control room, the
operator would be able to direct and monitor recovery efforts from
the control room that may be necessary to bring plant parameters
within required specifications.
If an operator remains in the control room during a tsunami and
becomes injured, that operator would be unable to direct and monitor
recovery efforts. Under this scenario, other plant personnel who
evacuated to higher ground onsite within the OCA would eventually
return to the plant, including the control room, and perform any
required recovery functions. Therefore, consequences of a tsunami
are not increased by not continually manning the control room during
the event.
2. Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
The proposed editorial change has no impact on accidents. The
following discussion applies to the proposed change related to
control room evacuation.
As discussed in the response to question 1 above, none of the
analyzed accidents require operator action to keep offsite
radiological doses well within regulatory limits. In addition,
allowing plant personnel to not continuously man the control room
after an emergency situation has occurred, has no impact on the
possibility of a new or different kind of accident from occurring.
If the plant is evacuated, no work activities will be performed in
the plant. With the plant in SAFSTOR and no work being performed,
there are no actions required to be taken by personnel manning the
control room.
3. Does the change involve a significant reduction in a margin
of safety?
Response: The proposed editorial change has no impact on margin
of safety. The following discussion applies to the proposed change
related to control room evacuation.
NRC SER Section 10.8, ``Accident Analysis Conclusions,''
summarizes the consequences from accidents in terms of offsite
radiological doses. SER Section 10.8 includes the statement, ``The
(NRC) staff has determined that offsite radiological consequences
due to a tsunami are within acceptable dose guideline values.'' As
discussed in the response to question 1 above, none of the analyzed
accidents require operator action to keep offsite radiological doses
well within regulatory limits. Therefore, temporarily not manning
the control room during an emergency will have no impact on the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based upon
the staff's review of the licensee's analyses as well as the staff's
own evaluation, the staff concludes 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: Richard F. Locke, Esquire, Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Claudia Craig.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: January 31, 2006.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.8.3.1, ``Onsite Power Distribution-
Operating,'' to extend the allowed outage time (AOT) for an inoperable
Class 1E vital 120-volt alternating current inverter. The TS currently
provides an AOT of 24 hours to restore an inoperable inverter. Based on
risk-informed assessment, the amendments would extend the AOT to 7
days.
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 formatting changes to TS 3.8.3.1 Action b and the
change to the AOT
[[Page 10078]]
for an inoperable inverter to be extended from 24 hours to 7 days do
not alter any plant equipment or operating practices in such a
manner that the probability of an accident is increased. The
proposed changes will not alter assumptions relative to the
mitigation of an accident or transient event.
An evaluation was performed to determine the risk significance
of the proposed change to the AOT. The risk evaluation concludes
that the [Delta]CDF [core damage frequency] and [Delta]LERF [large
early release frequency] associated with the proposed changes are
1.88E-07 and 2.05E-09, respectively, which are characterized as
``very small changes'' by RG [Regulatory Guide] 1.174. The ICCDP
[incremental conditional core damage probability] and ICLERP
[incremental conditional large early release probability] associated
with the proposed change are 3.63E-07 and 1.08E-08, respectively,
which are within the acceptance criteria in RG 1.177. 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 a physical alteration of the
plant (no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. Therefore,
the proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel and fuel cladding, reactor
coolant pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed change to TS
3.8.3.1 to allow the AOT for an inoperable inverter to be extended
from 24 hours to 7 days has been evaluated for its effect on plant
safety. The risk-informed evaluation concludes that the [Delta]CDF
and [Delta]LERF associated with the proposed change are 1.88E-07 and
2.05E-09, respectively, which are characterized as ``very small
changes'' by RG 1.174. The ICCDP and ICLERP associated with the
proposed change are 3.63E-07 and 1.08E-08, respectively, which are
within the acceptance criteria in RG 1.177. The proposed changes to
the formatting of TS 3.8.3.1 Action b are administrative only and
have no impact on margin of safety. Therefore, the proposed 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 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: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: David Terao.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant, Unit 1 (WBN) Rhea County, Tennessee
Date of amendment request: December 14, 2005 (TS-05-07).
Description of amendment request: The proposed amendment would
revise Technical Specification Section 5.7.2.19, ``Containment Leakage
Rate Testing Program,'' to allow a one time, 5-year extension to the
current 10-year test interval for the performance-based leakage rate
test program for 10 CFR Part 50, Appendix J, Type A tests.
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 for extending Type A test frequency does not
significantly increase the probability of an accident previously
evaluated since the change is not a modification to plant systems,
nor a change to plant operation that could initiate an accident.
TVA performed an evaluation of the risk significance for the
proposed increase to the WBN Unit 1 Type A test frequency. The
results of the TVA risk evaluation indicates that the increase in
Large Early Release Frequency (LERF) remains below the level of risk
significance defined in the NRC Regulatory Guide 1.174, ``An
Approach for Using Probabilistic Risk Assessment In Risk-Informed
Decisions On Plant-Specific Changes to the Licensing Basis.'' TVA's
evaluation indicates that the calculated increase in frequency for
all releases (small, large, early and late) and the increase in
radiation dose to the population are also non-risk significant.
The proposed test interval extension does not involve a
significant increase in the consequences of an accident. Research
documented in NUREG-1493, ``Performance-Based Containment Leakage-
Test Program,'' determined that generically, very few potential
containment leakage paths fail to be identified by Type A tests. An
analysis of 144 Type A test results, including 23 failures, found
that no failures were due to containment liner breach. The NUREG
concluded that reducing the Type A test frequency to once per 20
years would lead to an imperceptible increase in risk. Furthermore,
the NUREG concluded that Type B and C testing provides assurance
that containment leakage from penetration leak paths (i.e., valves,
flanges, containment air-locks) identify any leakage that would
otherwise be detected by the Type A tests.
In addition to the NUREG conclusions, TVA's American Society of
Mechanical Engineers (ASME) IWE program performs containment
inspections in order to detect evidence of degradation that may
either affect the containment structural integrity or leak
tightness.
Therefore, the proposed extension of the Type A test interval
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 to extend the Type A test interval does not
create the possibility of a new or different type of accident
because there are no physical changes made to the plant or plant
equipment governing normal plant operation. There are no changes to
the operation of the plant that would introduce a new failure mode
creating the possibility of a new or different kind of accident.
Therefore, the proposed extension does not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to extend the Type A test interval will not
significantly reduce the margin of safety. A generic study
documented in NUREG-1493 indicates that extending the Type A leak
test interval to 20 years would result in an imperceptible increase
in risk to the public. The NUREG also found that, generically, the
containment leakage rate contributes a very small amount to the
individual risk and that the decrease in the Type A test frequency
would have a minimal effect on risk because most potential leakage
paths are detected by Type C testing.
Previous Type A leakage tests conducted on WBN Unit 1 indicate
that leakage from containment have been less than the 10 CFR 50,
Appendix J leakage limit of 1.0 La. A review of the
previous Type A test results indicate a stable trend with an
increase of less than 15 percent of La, well below the
1.0 La leakage limit.
Therefore, these test results, in conjunction with the research
findings from NUREG-1493, provide assurance that the proposed
extension to the Type A test interval 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
[[Page 10079]]
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: August 26, 2005, as supplemented by
letter dated December 16, 2005.
Description of amendment request: The amendment would authorize
changes to the Final Safety Analysis Report (FSAR) for the Callaway
Plant, Unit 1, that would revise the methodology for the reactor
coolant system (RCS) leak detection instrumentation. This revision
would clarify the requirements of the containment atmosphere gaseous
radioactivity monitor with regard to the RCS leak detection capability
and would justify that the monitor can be considered operable in
compliance with Limiting Condition for Operation 3.4.15, in Technical
Specification (TS) 3.4.15, ``RCS Leakage Detection Instrumentation,''
during all applicable reactor modes. There are no proposed changes to
the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change has been evaluated and determined to not
increase the probability or consequences of an accident previously
evaluated. The proposed change does not make hardware changes and
does not alter the configuration of any plant system, structure, or
component (SSC). The proposed change only clarifies the design and
OPERABILITY requirements for the containment atmosphere gaseous
radioactivity monitor[s] and identifies the capabilities of the
containment atmosphere gaseous radioactivity monitors at low RCS
[radio]activity levels. The containment radiation monitors are not
initiators of any accident; therefore, the probability of occurrence
of an accident is not increased. The FSAR and TS will continue to
require diverse means of [RCS] leakage detection equipment, thus
ensuring that leakage due to cracks [in the RCS] would continue to
be identified prior to propagating to the point of a [RCS] pipe
break. Therefore, the consequences of an accident [previously
evaluated] are not increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve the use or installation of
new equipment and the currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bas[i]s [for the Callaway Plant]. The proposed
change does not affect any SSC associated with an accident
initiator. Based on this evaluation, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not alter any RCS leakage detection
components. The proposed change only clarifies the design and
OPERABILITY requirements for the containment atmosphere gaseous
radioactivity monitor[s] and identifies the capabilities of the
containment atmosphere gaseous radioactivity monitors at low RCS
[radio]activity levels. This change is required since the level of
radioactivity in the Callaway Plant reactor coolant has become much
lower than what was assumed in the FSAR [when the plant was
licensed] and the gaseous channel [(monitor)] can no longer promptly
detect a small RCS leak under all operating conditions. The proposed
amendment continues to require diverse means of [RCS] leakage
detection equipment with [the] capability to promptly detect RCS
leakage. Although not required by TS, additional diverse means of
leakage detection capability are available as described in the FSAR
Section 5.2.5. Early detection of [RCS] leakage, as the potential
indicator of a crack(s) in the RCS pressure boundary, will thus
continue to be in place so that such a condition is known and
appropriate actions taken well before any such crack would propagate
to a more severe condition. Based on this evaluation, 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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 1, 2006.
Description of amendment request: The amendment would revise the
Inservice Testing Program in Section 5.5.8 of the Administrative
Controls, Programs and Manuals, section of the Technical Specifications
(TSs). The licensee is adopting NRC-approved Technical Specification
Task Force (TSTF) 479, Revision 0, ``Changes to Reflect Revision of 10
CFR 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)
regarding the inservice testing of pumps and valves. The proposed
change incorporates revisions to the ASME [American Society of
Mechanical Engineers] Code [for Operation and Maintenance of Nuclear
Power Plants] that result in a net improvement in the measures for
testing pumps and valves.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. They do not involve the addition or removal of any
equipment, or any design changes to the facility. Therefore, the
proposed change does not represent a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)
regarding the inservice testing of pumps and valves. The proposed
change incorporates revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves.
The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure. Therefore, this proposed change
does not create the possibility of an accident of a different kind
than previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The proposed change revises TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)
regarding the inservice testing of pumps and valves. The proposed
change incorporates revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves. The safety
function of the affected pumps and valves will be
[[Page 10080]]
maintained. Therefore, this proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
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: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 7, 2006.
Description of amendment request: The amendment would add
Surveillance Requirement (SR) 3.3.1.16, to verify the reactor trip
system response time, to Function 3.a, power range neutron flux--high
positive rate trip function, in Table 3.3.1-1, ``Reactor Trip System
Instrumentation,'' of the Technical Specifications (TSs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Overall protection system performance will remain within the
bounds of the accident analysis since there are no hardware changes.
The design of the Reactor Trip System (RTS) instrumentation,
specifically the positive [neutron] flux rate trip (PFRT) function,
will be unaffected. The reactor protection system will continue to
function in a manner consistent with the plant design basis. All
design, material, and construction standards that were applicable
prior to the request [(i.e., this amendment application)] are
maintained.
The proposed change imposes additional surveillance requirements
to assure safety related structures, systems, and components are
verified to be consistent with the [plant] safety analysis and
licensing basis. In this specific case, a response time verification
requirement will be added to the PFRT Function [in TS Table 3.3.1-
1].
The proposed [change] will not modify any system interface. The
proposed [change] will not affect the probability of any event
initiators. There will be no degradation in the performance of or an
increase in the number of challenges imposed on safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance. The proposed [change] will not alter any
assumptions or change any mitigation actions in the radiological
consequence evaluations in the Updated Safety Analysis Report (USAR)
[for Wolf Creek Generating Station].
The proposed [change does] not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated or maintained. The proposed [change does] not
alter or prevent the ability of structures, systems, and components
(SSCs) from performing their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed [change does] not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. The proposed [change is] consistent with the safety
analysis assumptions and resultant consequences.
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.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. This change will not affect the normal method of plant
operation or change any operating parameters. No performance
requirements will be affected; however, the proposed change does
impose additional surveillance requirements. The additional
requirements are consistent with assumptions made in the safety
analysis and licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of [the change]. There will be no adverse effect or challenges
imposed on any safety-related system as a result of [the change].
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 does] not affect the acceptance criteria
for any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). 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. There will be no impact on the overpower limit, DNBR
[departure from nucleate boiling ratio] limit, FQ [heat
flux hot channel factor], F[utri]H [nuclear enthalpy rise hot
channel factor], LOCA PCT [loss-of-coolant accident peak cladding
temperature], peak local power density, or any other margin of
safety. The radiological dose consequence acceptance criteria listed
in the [NRC] Standard Review Plan [NUREG-0800] will continue to be
met.
The safety analysis limits assumed in the transient and accident
analyses are unchanged. None of the acceptance criteria for any
accident analysis is changed. The imposition of additional
surveillance requirements increases the margin of safety by assuring
that the affected safety analysis assumptions on equipment response
time are verified on a periodic frequency. 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: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
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.
[[Page 10081]]
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 Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: August 11, 2005.
Brief Description of amendments: The amendments revise Technical
Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing
Program,'' by removing an exception that allows for compensation of
flow meter instrument inaccuracies in accordance with ANSI/ANS-56.8-
1987 rather than ANSI/ANS-56.8-1994.
Date of issuance: February 8, 2006.
Effective date: Date of issuance to be implemented within 60 days.
Amendment Nos.: 238 and 266.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the TS.
Date of initial notice in Federal Register: September 13, 2005 (70
FR 54087).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 8, 2006.
No significant hazards consideration comments received: No.
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 application for amendment: June 8, 2005.
Brief description of amendment: The proposed changes would add
Limiting Condition for Operation 3.0.8 to address conditions where one
or more snubbers are unable to perform their associated support
function.
Date of issuance: February 13, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 245 and 229.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48203).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 13, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: March 7, 2005, as supplemented
by letter dated December 5, 2005.
Brief description of amendments: The amendments will add two
Nuclear Regulatory Commission (NRC) approved topical report references
to the list of analytical methods in Technical Specification 5.6.5,
``Core Operating Limits Report,'' that can be used to determine core
operating limits.
Date of issuance: February 1, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 174 and 160.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48205).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 1, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: December 17, 2004.
Brief description of amendments: The amendments revised the
Appendix B, Environmental Protection Plan (non-radiological), of the
Quad Cities Station Renewed Facility Operating Licenses.
Date of issuance: February 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 229 and 224.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Environmental Protection Plan.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19115).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 2, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County,
Pennsylvania
Date of application for amendment: April 13, 2005, as supplemented
by letters dated August 26, October 28 and 31, November 18, and
December 6 and 16, 2005.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to allow replacement of the BVPS-1 steam
generators (SGs). These changes include revising the fuel assembly-
specific departure from nucleate boiling ratios and correlations,
modifying the Overtemperature [Delta]T and Overpower [Delta]T
equations, revising the SG water level low-low and high-high setpoints,
revising the SG secondary side level in Modes 4 and 5, revising the SG
TSs to reflect the replacement SGs and remove TS requirements that are
no longer applicable to the new SGs, revising the required charging
pump discharge pressure for reactor coolant pump seal injection flow,
raising the accumulator pressure, and adding WCAP-14565-P-A (VIPRE) and
WCAP-15025-P-A (WRB-2M) Topical Reports to the list of NRC-approved
methodologies listed in TS 6.9.5. The amendment also approves an
expanded selective alternate source term methodology implementation in
accordance with Regulatory Guide 1.183, ``Alternate Radiological Source
Terms for Evaluating Design Basis Accidents at Nuclear Power
Reactors,'' and approves use of the 1979 ANS Decay Heat + 2[sigma]
model for mass and energy releases for a main steam line break outside
containment.
Date of issuance: February 9, 2005.
Effective date: As of its date of issuance and shall be implemented
prior to entry into Mode 4 upon startup from refueling outage 1R17
which begins on or about February 10, 2006.
Amendment No: 273.
[[Page 10082]]
Facility Operating License No. DPR-66: The Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 21, 2005 (70 FR
35737). The supplements dated August 26, October 28 and 31, November
18, and December 6 and 16, 2005, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: October 4, 2004, as
supplemented July 8, and November 14, 2005.
Brief description of amendments: These amendments approved
application of the Westinghouse best-estimate loss-of-coolant accident
(LOCA) analysis methodology to BVPS-1 and 2 for large-break LOCA
analysis.
Date of issuance: February 6, 2006.
Effective date: These license amendments are effective as of the
date of issuance and shall be implemented for BVPS-1, prior to Mode 4
entry during startup from refueling outage 1R17 which begins on or
about February 10, 2006, and for BVPS-2, prior to Mode 4 entry during
startup from refueling outage 2R12 which begins October 2006.
Amendment Nos.: 272 and 154.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70718). The supplements dated July 8, and November 14, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the Nuclear Regulatory Commission staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 6, 2006.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendment: August 10, 2005.
Brief description of amendment: The amendments deleted the power
range neutron flux high negative rate trip function from Table 3.3.1-1,
``Reactor Trip System Instrumentation.''
Date of issuance: February 10, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 293, 275.
Facility Operating License No. DPR-58: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72674). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 10, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: June 7, 2004, as supplemented by
letters dated February 18, May 20, June 16, July 8, August 3, September
23, and November 16, 2005, and February 6, 2006.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to reflect an expanded operating domain resulting
from the implementation of the Average Power Range Monitor, Rod Block
Monitor TSs/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA).
Date of issuance: February 8, 2006.
Effective date: As of the date of issuance, to be implemented
within 120 days.
Amendment No.: 163.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: September 14, 2004 (69
FR 55471). The supplements dated February 18, May 20, June 16, July 8,
August 3, September 23, and November 16, 2005, and February 6, 2006,
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 amendment is contained
in a Safety Evaluation dated February 8, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: February 1, 2005, supplemented
by letters dated February 22, September 16, December 2, 2005, and
January 5, 2006.
Brief description of amendments: The amendments revise the spent
fuel pool (SFP) criticality analysis methodology and technical
specifications governing the storage of irradiated fuel in the SFP. The
licensee's amendment request stated that subcritical conditions would
be maintained in the SFP under the revised technical specification
storage requirements.
Date of issuance: February 5, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 172, 162.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 2005, (70 FR
12748). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendments
is contained in a Safety Evaluation dated February 5, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: January 19, 2005, as supplemented on
June 9 (two letters) and November 18, 2005.
Brief Description of amendments: The amendment authorizes revision
of the Updated Final Safety Analysis Report (UFSAR) to reflect the
utilization of fire-rated electrical Mineral Insulated cables in lieu
of Appendix R, Section III.G.2 1-hour rated fire barriers.
Date of issuance: February 13, 2006.
Effective date: As of the date of issuance, to be incorporated into
the UFSAR at the time of its next update.
Amendment No.: 162.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendment
authorizes revision to the UFSAR.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21464). The supplemental letters provided clarifying information that
was within
[[Page 10083]]
the scope of the initial notice and did not change the initial proposed
no significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated February 13, 2006.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, 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, 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/
[[Page 10084]]
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 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 or 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).
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: February 5, 2006, as supplemented
February 5, 2006.
Description of amendment request: The amendment revised Technical
Specification 3.8.1, ``AC Sources--Operating,'' to extend the allowed
outage time for Emergency Diesel Generator 12 from seven days to 14
days for one specific incident.
Date of issuance: February 6, 2006.
Effective date: As of the date of issuance and shall be implemented
immediately.
Amendment No.: 171.
Facility Operating License No. 50-341: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated
February 6, 2006.
Attorney for licensee: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Branch Chief: Timothy J. Kobetz, Acting.
Dated at Rockville, Maryland, this 16th day of February, 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-1737 Filed 2-27-06; 8:45 am]
BILLING CODE 7590-01-P