[Federal Register Volume 70, Number 118 (Tuesday, June 21, 2005)]
[Notices]
[Pages 35735-35743]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-3138]
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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 May 26, 2005, to June 9, 2005. The last
biweekly notice was published on June 7, 2005 (70 FR 33210).
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.
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
[[Page 35736]]
with particular reference to the following general requirements: (1)
The name, address, and telephone number of the requestor or petitioner;
(2) the nature of the requestor's/petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
requestor's/petitioner's property, financial, or other interest in the
proceeding; and (4) the possible effect of any decision or order which
may be entered in the proceeding on the requestor's/petitioner's
interest. The petition must also set forth the specific contentions
which the petitioner/requestor seeks to have litigated at the
proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of amendment request: May 25, 2005.
Description of amendment request: The amendment would revise
Technical Specification Section 3.4.9, ``Pressurizer,'' to revise the
pressurizer water level limit during operation in Mode 3 (hot standby).
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.
Pressurizer water level is an assumed initial condition for
certain accident analyses. Plant initial conditions are not accident
initiators and do not have an effect on the probability of the
accident occurring. The proposed change only revises the specified
limit on water level in the pressurizer, so this change does not
affect accident probability.
Pressurizer water level is an assumed initial condition for
accidents such as LOCA [loss-of-coolant accident], loss-of-load and
loss-of-normal feedwater. The limiting accident analysis results
occur at full power conditions when the available core thermal power
is maximized. The proposed change does not affect the specified
pressurizer level limit at any power level from zero to full power.
That is, the pressurizer level limit is not being changed in Modes 1
and 2. The proposed change does revise the specified pressurizer
water level limit in Mode 3 (Hot Standby) but this does not affect
accident analysis results because the limiting analyses will remain
those that are postulated to occur in Mode 1 with the plant at full
power.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve physical changes to
existing plant equipment or the installation of any new equipment.
The design of the pressurizer, the pressurizer level control system
and the pressurizer safety valves is not being changed and the
ability of these systems, structures, and components to perform
their design or safety functions is not being affected. The proposed
change revises the specified limit on pressurizer water level in
Mode 3 (Hot Standby) to allow operators greater flexibility in
performing a plant cooldown. The method used in performing the plant
cooldown is not
[[Page 35737]]
being changed. This proposed change does not create new failure
modes or malfunctions of plant equipment nor is there a new credible
failure mechanism.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Pressurizer level is an initial condition assumed in certain
accident analyses involving an insurge in the pressurizer and an
increasing reactor coolant system (RCS) pressure. These analyses
demonstrate that the design pressure for the RCS is not exceeded for
the limiting analyses based on the plant at full power. The proposed
change does not affect the existing Technical Specification
requirement for Mode 1 (Power Operation) or Mode 2 (Plant Startup)
and therefore does not affect the assumptions or results of these
accident analyses. The margin for RCS design pressure demonstrated
by these analysis results is not being reduced. The proposed change
only applies to the pressurizer level limit in Mode 3 (Hot Standby)
when there is substantially lower thermal energy available to cause
rapid expansion of reactor coolant and an insurge to the
pressurizer. Protection of the RCS pressure boundary is still
maintained by the pressurizer safety valves, which are not being
modified by the proposed change in pressurizer water level.
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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 15, 2005.
Description of amendment request: A change is proposed to revise
the Waterford Steam Electric Station Unit 3 (Waterford 3) Technical
Specification (TS) Section 4.4.4.4 to modify the steam generator tube
inspection Acceptance Criteria for the ``Plugging or Repair Limit'' and
the ``Tube Inspection,'' as contained in the Waterford 3 TS
Surveillance Requirements (SR) 4.4.4.4.a.7 and 4.4.4.4.a.9,
respectively. The purpose of these changes is to define the depth of
the required tube inspections and to clarify the plugging criteria
within the tubesheet region.
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.
Conducting the rotating Plus Point probe inspections to a
minimum tubesheet length of 10.4 inches maintains the existing
design limits and does not increase the probability or consequences
of an accident involving tube burst or primary to secondary
accident-induced leakage, as previously analyzed in the Waterford 3
Final Safety Analysis Report. Also the NEI [Nuclear Energy
Institute] 97-06 structural integrity and accident induced leakage
of the steam generator tubes performance criteria will continue to
be satisfied.
Tube burst is precluded for a tube with defects within the
tubesheet region because of the constraint provided by the
tubesheet. As such, tube pullout resulting from the axial forces
induced by primary to secondary differential pressures would be a
prerequisite for tube burst to occur. Any degradation below C* is
shown by empirical test results and analyses to be acceptable,
thereby precluding an event with consequences similar to a
postulated tube rupture event. WCAP-16208-P has shown that tube
flaws below the C* length will not result in primary to secondary
leakage greater than 0.1 gpm [gallons per minute] per steam
generator. Inspection to the C* length will ensure that the
postulated accident induced leakage for events that involve a
faulted steam generator (e.g., a main steam line break (MSLB)) will
remain within both the current and proposed extended power uprate
(EPU) accident analyses of 720 gpd (0.5 gpm) and 540 gpd (0.375
gpm), respectively.
Therefore, the proposed change does not affect the probability
or consequences of any Waterford 3 analyzed accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Steam generator tube leakage and structural integrity will be
maintained during all plant conditions upon implementation of the
proposed inspection scope and plugging or repair limit changes to
the Waterford 3 Technical Specifications. These changes do not
introduce any new mechanisms that might result in a different kind
of accident from those previously evaluated. Even with the limiting
circumstances of a complete circumferential separation (360o through
wall crack) of all of the tubes below the C* length, tube pullout is
precluded and leakage is predicted to be maintained within both the
current and proposed extended power uprate (EPU) accident analyses
assumptions.
Therefore, the proposed change does not create the possibility
of a new or different type 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 inspection and plugging criteria will better assure
that steam generator tube performance is maintained within its
design basis and within the safety analysis assumptions. Operation
with potential tube degradation below the C* inspection length
within the tubesheet region of the steam generator tubing meets the
intent of the inspection guidance of RG 1.83, Inservice Inspection
of Pressurized Water Reactor Steam Generator Tubes, the requirements
of General Design Criteria 14, 30 and 32 of 10 CFR 50, and the
recommendations of NEI-97-06, Steam Generator Program Guidelines.
The total leakage from an undetected flaw population below the C*
inspection length under postulated accident conditions is accounted
for to assure that the leakage criterion is met and bounded by both
the current and the proposed EPU accident analyses assumptions.
Adequate margin remains for other possible steam generator tube leak
sources.
The proposed changes also maintain the structural and accident-
induced leakage integrity of the steam generator tubes as required
by NEI 97-06 and the plant design basis.
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: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: David Terao.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County,
Pennsylvania
Date of amendment request: April 11, 2005.
Description of amendment request: The proposed amendment would
revise the BVPS-1 Technical Specifications (TSs) to permit operation
with replacement Model 54F steam generators (SGs) installed. These
include changes to reactor core safety limits, reactor trip system and
engineered safety features actuation system setpoints, and other safety
analysis inputs related to the proposed new model 54F steam generators
as well as changes to steam generator limiting conditions for operation
and surveillance requirements. These proposed TS changes were
originally submitted as part of the licensee's extended power uprate
application,
[[Page 35738]]
dated October 4, 2004, however, delays in the review of that
application have required the licensee to separately request these
proposed TS changes in order to support SG replacement during and
startup from the BVPS-1 2006 refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed changes will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The safety and radiological dose consequence analyses confirmed
that safety analysis and dose consequence analysis acceptance
criteria will be satisfied for the Model 54F BVPS Unit No. 1
replacement steam generators, including changes to reactor core
safety limits, reactor trip system (RTS) and engineered safety
features actuation system (ESFAS) setpoints, and other safety
analysis inputs related to the proposed changes. The analyses are
conservative and bounding with respect to operation with RSGs
[replacement steam generators] at the current licensed maximum power
level.
For the purpose of this evaluation, the proposed changes to
Technical Specifications 3.4.1.3, Reactor Coolant system Shutdown,
and 3.4.5, Steam Generators, which will directly address the new
Unit No. 1 replacement steam generators (RSG) can be grouped in the
following areas:
(a) The first area of change is to remove the references to
repair of tubes by sleeving since they are not applicable to the RSG
tubes.
The accidents of interest are [steam generator] tube rupture and
steam line break. A reduction in tube integrity could increase the
possibility of a tube rupture accident and could increase the
consequences of a steam line break. The tubing in the RSGs is
designed and evaluated consistent with the margins of safety
specified in the ASME Code [American Society of Mechanical
Engineers, Boiler and Pressure Vessel Code], Section III. The
program for periodic inservice inspection provides sufficient time
to take proper and timely corrective action if tube degradation is
present. The basis for the 40% through wall plugging limit is
applicable to the RSGs just as it was to the original steam
generators (OSG). An analysis has been performed consistent with the
guidance in Draft Regulatory Guide 1.121 to justify the
applicability of the 40% through wall plugging limit. As a result,
there is no reduction in tube integrity for the RSGs.
Elimination of the repair option and the associated references
to repair of the OSG tubes is an administrative adjustment since the
sleeve design is not applicable to the RSGs. The elimination of the
repair option does not alter the requirements for inservice
inspection or reduce the plugging limit for the RSG tubes.
(b) The second area of change is to remove the references to
voltage-based repair criteria on tube-to-tube support plate
intersections since they are not applicable to the RSG tubes.
Elimination of the repair option and the associated repair of
the OSG tubes is an administrative adjustment since the voltage
based repair criteria is not applicable to the RSGs. The elimination
of the repair option does not alter the requirements for inservice
inspection or reduce the plugging limit for the RSG tubes.
(c) The third area of change is to update the wording and
content of the TS to provide clarification and to incorporate
wording enhancements consistent with the updates made to the subject
TS for several other plants that have replaced steam generators.
Since the RSGs will be subjected to a preservice inspection prior to
installation, there is no need to perform inservice inspection
following installation.
The changes to update the wording and content of the TS to
provide clarification and to incorporate wording enhancements are
administrative changes that provide clarifications. These changes do
not alter the requirements for inservice inspection or the plugging
limit for the tubes.
(d) The fourth area of change is to revise the steam generator
water levels.
The proposed steam generator water level setpoint changes do not
impact the initiation of accidents; therefore, they do not involve
an increase in the probability of an accident previously evaluated.
The proposed changes do impact the safety analyses for accidents
that credit the applicable trips and associated system actions;
however, they do not alter these accidents or the associated
accident acceptance criteria. The safety analyses for these
accidents have been performed at 2900 MWt [megawatts thermal] (which
is conservative and bounding for the current licensed power level of
2689 MWt) and show acceptable results. Therefore, the proposed
changes do not involve a significant increase in the consequences of
an accident previously evaluated.
The proposed change to steam generator water level used to
verify steam generator operability in Modes 4 and 5, i.e., TS
3.4.1.3, does not impact the initiation of accidents; therefore, it
does not involve an increase in the probability of an accident
previously evaluated. The proposed change does not alter the safety
analyses for accidents or the associated accident acceptance
criteria. Therefore, the proposed change does not involve a
significant increase in the consequences of an accident previously
evaluated.
The proposed changes, due to the replacement steam generators,
do not alter the requirements for tube inspection, tube integrity,
or tube plugging limit, therefore they do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Use of the VIPRE computer code and the WRB-2M correlation at
BVPS for departure from nucleate boiling (DNB) analysis for those
Updated Final Safety Analysis Report (UFSAR) transients and
accidents for which DNB might be a concern will not involve a
significant increase in the probability or consequences of an
accident previously evaluated for the following reasons. The code
and correlation are evaluation tools that are independent of the
probability of an accident. Use of the code and correlation
establish DNB limits such that core damage will not occur during
postulated design basis accidents. Thus, use of the code and
correlation will not involve a significant increase in the
consequences of an accident previously evaluated.
Use of the 1979 ANS [American Nuclear Society] Decay Heat +
2[sigma] 4 model for MSLB [main steam line break] outside
containment M&E [mass and energy] releases will not have a
significant increase in the probability or consequences of an
accident previously evaluated because the model is not an accident
initiator.
The remaining changes, which include the changes to the
Overtemperature [Delta]T and Overpower [Delta]T equations, the
change to the charging pump discharge pressure, and the additions of
WCAP-14565-P-A and WCAP-15025-P-A to the list of NRC approved
methodologies in TS 6.9.5, will not involve a significant increase
in the probability or consequences of an accident previously
evaluated because none of the changes are accident initiators.
The RSG radiological analysis reflects an expansion of the
selective application of the AST methodology and incorporation of
the ARCHON96 methodology for on-site atmospheric dispersion factors.
The radiological analysis concludes that normal operation of the
BVPS Unit No. 1 with the RSGs with an atmospheric containment will
not impact the unit's compliance with the normal operation operator
exposure limits set forth in 10 CFR 20 [Title 10 of the Code of
Federal Regulations, Part 20], or the public exposure limits set
forth in 10 CFR 20, 10 CFR 50, Appendix I and 40 CFR 190, or with
the post-accident exposure limits set forth by 10 CFR 100 or 10 CFR
50.67, as supplemented by Regulatory Guide 1.183, for the plant
operator and the public.
The effects on accident radiation dose considered the
replacement of the Unit No. 1 steam generators, a core power level
to 2900 MWt, incorporation of the ARCHON96 methodology and the
expansion of the selective implementation of the AST methodology.
None of these changes are initiators of any design basis accident or
event, and therefore, will not increase the probability of any
accident previously evaluated. The probability of any evaluated
accident or event is independent of these changes.
These proposed changes required alteration of some assumptions
previously made in the radiological consequence evaluations. The
assumption alterations were necessary to reflect the replacement
steam generators for Unit No. 1 and the incorporation of the
ARCHON96 and AST methodologies. These changes were evaluated for
their effect on accident dose consequences. The updated dose
consequence analyses demonstrate compliance with the limits set
forth for AST
[[Page 35739]]
applications in 10 CFR 50.67, as supplemented by Regulatory Guide
1.183 or 10 CFR part 100.
Therefore, in conclusion, none of the proposed changes involve a
significant increase in the probability of an accident previously
evaluated, and the dose consequences remain within the allowable
limits set forth for AST applications in 10 CFR 50.67, as
supplemented by Regulatory Guide 1.183 or 10 CFR part 100.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The areas of change described previously for the Unit No. 1 RSGs
do not adversely affect the design or function of any other safety-
related component. With respect to postulated accident conditions,
the OSGs and the RSGs are the same. There is no mechanism to create
a new or different kind of accident for the RSGs by eliminating
repair criteria or by clarifying the applicability of inservice
inspection requirements because a baseline of tube conditions is
established and plugging limits are maintained to ensure that
defective tubes are identified and removed from service.
The proposed changes to steam generator water level setpoints,
and the steam generator water level used to verify steam generator
operability in Modes 4 and 5 do not impact the initiation of
accidents. They do not alter the accidents that credit the
associated trips or accident acceptance criteria. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes do not alter the requirements for tube
inspection, tube integrity, or tube plugging limit; therefore, they
do not create the possibility of a new or different kind of accident
from any previously evaluated.
Use of the VIPRE computer code and WRB-2M correlation at BVPS
will not create the possibility of a new or different kind of
accident from any accident previously evaluated because the code and
correlation are evaluation tools. They are not accident initiators.
Thus, their use cannot create a new or different kind of accident.
Use of the 1979 ANS Decay Heat + 2[sigma] model for MSLB outside
containment M&E releases will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the model does not alter how any equipment is operated.
The remaining changes, which include the changes to the
Overtemperature [Delta]T and Overpower [Delta]T equations, the
change to the charging pump discharge pressure, and the additions of
WCAP-14565-P-A and WCAP-15025-P-A to the list of NRC approved
methodologies in TS 6.9.5, will not create the possibility of a new
or different kind of accident from any accident previously evaluated
because these changes do not alter how any equipment is operated.
The radiological changes will not create the possibility of a
new or different kind of accident from any previously evaluated
because they do not affect how components or systems are operated,
nor do they create new components or systems failure modes.
Therefore, in conclusion, none of the proposed changes create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed changes will not involve a
significant reduction in a margin of safety.
The steam generator tube integrity provides the margin of
safety. The tubing in the RSGs is designed and evaluated consistent
with the margins of safety specified in the ASME Code, Section III.
The program for periodic inservice inspection provides sufficient
time to take proper and timely corrective action if tube degradation
is present. The basis for the 40% through wall plugging limit is
applicable to the RSGs just as it was to the OSGs. A Regulatory
Guide 1.121 analysis was performed to confirm the applicability of
the 40% through wall plugging limit. As a result, there is no
reduction in tube integrity for the RSGs.
The proposed changes to steam generator water level setpoints do
not alter the reactor trip system/engineered safety features
actuation system setpoint analysis methodology, or the associated
accident analysis methodology or acceptance criteria. The safety
analyses for these accidents have been performed at a power level of
2900 MWt ( which is conservative and bounding for the current
licensed power level of 2689 MWt) and show acceptable results.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The proposed change to the steam generator water level used to
verify steam generator operability in Modes 4 and 5 does not alter
the steam generator water level uncertainty and setpoint analysis
methodology or the associated natural circulation analysis
methodology or acceptance criteria. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
The proposed changes to update the wording and content of the TS
to provide clarification and to incorporate wording enhancements are
administrative changes that provide clarifications.
The proposed changes do not alter the requirements for tube
integrity, tube inspection or tube plugging limit; therefore, they
do not involve a significant reduction in a margin of safety.
Use of the VIPRE computer code and the WRB-2M correlation at
BVPS will not involve a significant reduction in a margin of safety
because the code and correlation are used to establish a margin of
safety previously approved by the NRC such that core damage will not
occur.
Use of the 1979 ANS Decay Heat + 2[sigma] model for MSLB outside
containment M&E releases will not involve a significant reduction in
a margin of safety because the results of the subject accident have
been shown to produce acceptable results.
The remaining changes, which include changes to the
Overtemperature [sigma]T and Overpower [sigma]T equations, the
change to the charging pump discharge pressure, and the additions of
WCAP-14565-P-A and WCAP-15025-P-A to the list of NRC approved
methodologies in TS 6.9.5, will not involve a significant reduction
in a margin of safety because they are being made to maintain the
existing margin of safety.
The radiological changes will not involve a significant
reduction in a margin of safety because BVPS compliance with the
limits set forth in 10 CFR 20, 10 CFR 50, Appendix I, 40 CFR 190, 10
CFR 100 and 10 CFR 50.67, as supplemented by Regulatory Guide 1.183,
will be maintained following approval of the requested changes.
A FENOC assessment of the cumulative effect of the proposed
changes provides [a] reasonable expectation that collectively they
will not result in a significant reduction in the overall margin of
safety. The results of the analyses demonstrate that the applicable
design and safety criteria and regulatory requirements will continue
to be met following approval of the proposed changes.
Therefore, in conclusion, none of the propose changes 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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
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 amendment request: June 1, 2005.
Description of amendment request: The amendments proposed by
Southern Nuclear Operating Company (SNC) would revise the Technical
Specifications (TS) to replace the previous TS requirement to implement
a Containment Tendon Surveillance Program based on Regulatory Guide
1.35, Rev. 2, ``Inservice Inspection of Ungrouted Tendons in
Prestressed Concrete Containment Structures,'' with a Containment
Inspection Program that complies with the current requirements of Title
10 of the Code of Federal Regulations (10 CFR) Section 50.55a, ``Codes
and Standards,'' in order to reflect the latest requirements for tendon
surveillance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 35740]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change replaces the previous TS requirement to
implement a Containment Tendon Surveillance Program based on
Regulatory Guide 1.35, Rev. 2, with a Containment Inspection Program
that complies with the current requirements of 10 CFR 50.55a. This
regulation requires licensees to implement a Containment Inspection
Program in compliance with the 1992 Edition with the 1992 Addenda of
Subsection IWE, ``Requirements for Class MC and Metallic Liners of
Class CC Components of Light-Water Cooled Plants,'' and with
Subsection IWL, ``Requirements for Class CC Concrete Components of
Light-Water Cooled Plants,'' of Section XI, Division 1, of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code) with additional modifications and limitations as
stated in 10 CFR 50.55a(b)(2)(ix). SNC has implemented a Containment
Inspection Program that complies with the regulatory requirements.
This proposed TS amendment is requested to update the TS to the
latest 10 CFR 50.55a regulatory requirements.
In addition, reporting requirements that are redundant to
existing regulations are deleted, minor editorial changes are made,
and the applicability of [Surveillance Requirement] SR 3.0.2 to the
tendon surveillance program is deleted since surveillance
frequencies and associated extensions are specified in ASME Section
XI, Subsection IWL.
By complying with the regulatory requirements described in 10
CFR 50.55a, the probability of a loss of containment structural
integrity is maintained as low as reasonably achievable. Maintaining
containment structural integrity as described in the revised
Containment Inspection Program does not impact the operation of the
reactor coolant system (RCS), containment spray (CS) system, or
emergency core cooling system (ECCS). The Containment Inspection
Program ensures that the containment will function as designed to
provide an acceptable barrier to release of radioactive materials to
the environment. 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 impact any accident initiators or analyzed
events, nor does it impact the types or amounts of radioactive
effluent that may be released offsite. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Maintaining containment structural integrity does not impact the
operation of the RCS, CS system, or ECCS. The proposed change does
not involve a modification to the physical configuration of the
plant or a change in the methods governing normal plant operation.
The proposed change does not introduce a new accident initiator,
accident precursor, or malfunction mechanism. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
By complying with the regulatory requirements described in 10
CFR 50.55a, the probability of a loss of containment structural
integrity is maintained as low as reasonably achievable. The
Containment Inspection Program ensures that the containment will
function as designed to provide an acceptable barrier to release of
radioactive materials to the environment. The proposed change does
not adversely affect plant operation or existing safety analyses.
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: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: Evangelos C. Marinos.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: May 26, 2005.
Description of amendment request: The amendment would change
Technical Specification (TS) 3.7.2, ``Main Steam Isolation Valves
(MSIVs),'' by adding the MSIV actuator trains to (1) the limiting
condition for operation (LCO) and (2) the conditions, required actions,
and completion times for the LCO. The existing conditions and required
actions in TS 3.7.2 are renumbered to account for the new conditions
and required actions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated.
Response: No.
The proposed changes to incorporate requirements for the MSIV
actuator trains do not involve any design or physical changes to the
facility, including the MSIVs and actuator trains themselves. The
design and functional performance requirements, operational
characteristics, and reliability of the MSIVs and actuator trains
are thus unchanged. There is therefore no impact on the design
safety function of the MSIVs to close (as an accident mitigator),
nor is there any change with respect to inadvertent closure of an
MSIV (as a potential transient initiator). Since no failure mode or
initiating condition that could cause an accident (including any
plant transient) evaluated per the FSAR [Callaway Final Safety
Analysis Report]-described safety analyses is created or affected,
the [proposed] change[s] cannot involve a significant increase in
the probability of an accident previously evaluated.
With regard to the consequences of an accident and the equipment
required for mitigation of the accident, the proposed changes
involve no design or physical changes to the MSIVs or any other
equipment required for accident mitigation. With respect to [the]
MSIV actuator train allowed outage times [(i.e., completion times)],
the consequences of an accident are independent of equipment allowed
outage times as long [as] adequate equipment availability is
maintained. The proposed MSIV actuator train allowed outage times
take into account the redundancy of the MSIV actuator trains and are
limited in extent consistent with other allowed outage times
specified in the Technical Specifications. Adequate equipment (MSIV)
availability would therefore continue to be required by the
Technical Specifications. On this basis, the consequences of
applicable, analyzed accidents (such as a main steam line break) are
not significantly impacted by the proposed changes. Based on all of
the above, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated.
Response: No.
None of the proposed changes, i.e., the addition of Conditions,
Required Actions and Completion Times [and addition to the LCO] to
[the] Technical Specifications for the MSIV actuator trains, involve
a change in the design, configuration, or operational
characteristics of the plant. No physical alteration of the plant is
involved, as no new or different type of equipment is to be
installed. The proposed changes do not alter any assumptions made in
the safety analyses, nor do they involve any changes to plant
procedures for ensuring that the plant is operated within analyzed
limits. As such, no new failure modes or mechanisms that could cause
a new or different kind of accident from any previously evaluated
are being introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
[[Page 35741]]
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety.
Response: No.
The proposed addition of Conditions, Required Actions and
Completion Times [and proposed addition to the LCO] to the Technical
Specifications for the MSIV actuator trains does not alter the
manner in which safety limits or limiting safety system settings are
determined. [There are no proposed changes to safety limits or
limiting safety system settings.] No changes to instrument/system
actuation setpoints are involved. The safety analysis acceptance
criteria are not impacted by [these proposed] change[s], and the
proposed change[s] will not permit plant operation in a
configuration outside the design basis.
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 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 Section Chief: Robert A. Gramm.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: October 21, 2004.
Description of amendment request: The amendment deletes the
Technical Specification (TS) requirements to submit monthly operating
reports and annual occupational radiation exposure reports. The change
is consistent with Revision 1 of the Nuclear Regulatory Commission
approved Technical Specifications Task Force (TSTF) Change Traveler,
TSTF-369, ``Elimination of Requirements for Monthly Operating Reports
and Occupational Radiation Exposure Reports.'' This TS improvement was
published in the Federal Register (69 FR 35067) on June 23, 2004, as
part of the Consolidated Line Item Improvement Process.
Date of issuance: June 8, 2005.
Effective date: June 8, 2005.
Amendment No.: 254.
Facility Operating License No. DPR-16: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. April 8, 2005 (70 FR 18056). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. Comments received from the State of New Jersey are
discussed in Section 7.0 of the related safety evaluation. The notice
also provided an opportunity to request a hearing by June 7, 2005, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated June 8, 2005.
Attorney for licensee: Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Richard J. Laufer.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: September 7, 2004.
Brief description of amendment: The amendment revised the required
frequency of quench and recirculation spray nozzle surveillances from
once every 10 years to ``following maintenance which could result in
nozzle blockage.'' The change also revised wording to correct grammar.
Date of issuance: May 31, 2005.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment No.: 222.
Facility Operating License No. NPF-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70715).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 31, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: June 3, 2003, as supplemented
by letter dated January 18 and May 10, 2005.
Brief description of amendments: The amendments would add a note to
Limiting Condition of Operation 3.7.11, ''Auxiliary Building Filtered
Ventilation Exhaust System (ABFVES),'' that would allow the Auxiliary
Building pressure boundary to be opened intermittently under
administrative control.
Date of issuance: June 2, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 229 and 211.
[[Page 35742]]
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 16, 2004 (69 FR
12365). The supplements dated January 18 and May 10, 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 amendments is contained
in a Safety Evaluation dated June 2, 2005.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 20, 2004.
Brief description of amendment: The amendment deletes the
requirements related to monthly operating reports and occupational
radiation exposure reports.
Date of issuance: May 25, 2005.
Effective date: As of the date of issuance, and shall be
implemented 90 days from the date of issuance.
Amendment No.: 145.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9990).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 25, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: December 7, 2004.
Brief description of amendment: This amendment revised the
Technical Specifications (TSs) by removing the surveillance requirement
(SR) for testing the setting of the standby liquid control system
pressure relief valves. Also, the SR for the recirculation pump
discharge valves was revised to remove stroke time specifications.
Date of Issuance: June 1, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 224.
Facility Operating License No. DPR-28: The amendment revised the
TSs.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2889).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated June 1, 2005.
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 amendments: April 6, 2004, as supplemented
by four letters dated April 15, 2005.
Brief description of amendments: The amendments convert the current
Technical Specifications (CTS) to the improved Technical Specifications
(ITS) and relocate license conditions to the ITS or other license
controlled documents. The ITS are based on NUREG-1431, ``Standard
Technical Specifications, Westinghouse Plants,'' dated April 30, 2001,
and guidance provided in the Commission's Final Policy Statement, ``The
U.S. Nuclear Regulatory Commission Final Policy Statement on Technical
Specifications (TSs) Improvements for Nuclear Power Reactors,''
published on July 22, 1993 (58 FR 39132), and 10 CFR Part 50.36,
``TSs.'' The overall objective of the proposed amendments was to
rewrite, reformat, and streamline the CTS to improve plant safety and
the understanding of the bases underlying the TSs.
Date of issuance: June 1, 2005.
Effective date: As of the date of issuance and shall be implemented
by October 30, 2005.
Amendment Nos.: 287, 269.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the TSs.
Date of initial notice in Federal Register: September 29, 2004 (69
FR 58205). 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 June 1, 2005.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: October 22, 2004.
Brief description of amendment: The amendment deleted Sections 5.3,
``Reactor Vessel,'' 5.4, ``Containment,'' and 5.6, ``Seismic Design,''
relocating all information, which pertains to design details, to the
Updated Final Safety Analysis Report.
Date of issuance: June 6, 2005.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 189.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70719).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 6, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: November 5, 2003, as
supplemented by letter dated April 22, 2004.
Brief description of amendments: The amendments revised the Point
Beach Nuclear Plant (PBNP), Units 1 and 2, Updated Final Safety
Analysis Report [UFSAR] to reflect the Commission staff's approval of
the WCAP-14439-P, Revision 2 analysis entitled, ``Technical
Justification for Eliminating Large Primary Loop Pipe Rupture as the
Structural Design Basis for the Point Beach Nuclear Plant Units 1 and 2
for the Power Uprate and License Renewal Program.''
Date of issuance: June 6, 2005.
Effective date: As of the date of issuance and shall be implemented
with the next update of the UFSAR in accordance with 10 CFR 50.71(e).
Amendment Nos.: 219, 224.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the License.
Date of initial notice in Federal Register: February 7, 2005 (70 FR
6466). The supplement dated April 22, 2004, provided clarifying
information that did not change the scope of the amendment, application
nor the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 6, 2005.
No significant hazards consideration comments received: No.
[[Page 35743]]
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: October 15, 2004.
Brief description of amendments: The amendments revised Technical
Specifications related to the reactor coolant pump flywheel inspection
program by increasing the inspection interval to 20 years.
Date of issuance: June 6, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 218, 223.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15945).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 6, 2005.
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: October 15, 2004.
Brief description of amendments: The amendments revise Technical
Specifications related to the reactor coolant pump flywheel inspection
program by increasing the inspection interval to 20 years.
Date of issuance: June 7, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 170, 160.
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 Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 7, 2005.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: September 23, 2004, and its
supplements dated December 21, 2004, and April 7, 2005.
Brief description of amendments: The amendments increase the
current minimum emergency diesel generator fuel oil inventory required
to be maintained onsite to support the use of low-sulfur fuel oil
required by California Air Resources Board.
Date of issuance: May 25, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 90 days from the date of issuance.
Amendment Nos.: Unit 1--181; Unit 2--183.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
402). The December 21, 2004, and April 7, 2005, supplemental letters
provided additional clarifying information, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 25, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: June 5, 2003, as supplemented
by letters dated June 3 and October 26, 2004.
Brief description of amendments: The amendments authorize changes
to the Updated Final Safety Analysis Report (UFSAR) for both units, to
acknowledge credit for possible operator action to ensure that the
containment design pressure is not exceeded in the event of a high
energy line break inside containment with a consequential failure of
the station control and service air system inside containment.
Date of issuance: May 24, 2005.
Effective date: As of the date of issuance and shall be implemented
as part of the next UFSAR update made in accordance with 10 CFR
50.71(e).
Amendment Nos.: 302 and 292.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
authorize changes to the UFSAR.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37584). The supplemental letters provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 24, 2005.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: October 27, 2004.
Brief description of amendment: The amendment revised Technical
Specification 3.7.3, ``Main Feedwater Isolation Valves (MFIVs),'' to
add the main feedwater regulating valves (MFRVs) and the associated
MFRV bypass valves (MFRVBVs). In addition, the allowed outage time, or
completion time, for inoperable MFIVs is extended.
Date of issuance: May 31, 2005.
Effective date: This amendment is effective as of its date of
issuance, and shall be implemented prior to entry into Mode 3 in the
restart from the upcoming Refueling Outage 14 (fall 2005).
Amendment No.: 167.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70722).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 31, 2005.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 10th day of June, 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. E5-3138 Filed 6-20-05; 8:45 am]
BILLING CODE 7590-01-P