[Federal Register Volume 69, Number 244 (Tuesday, December 21, 2004)]
[Notices]
[Pages 76486-76498]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-27614]
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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 November 25, 2004, through December 9, 2004.
The last biweekly notice was published on December 7, 2004 (69 FR
70712).
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 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
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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].
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: June 10, 2004.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.6.3, ``Containment Isolation Valves,''
to allow the surveillance frequencies for leakage rate testing to be
specified in the Catawba Nuclear Station Containment Leak Rate Testing
Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
First Standard
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No.
This amendment will not change any previously evaluated
accidents such as the postulated ``Fuel Handling Accident (FHA) in
Containment''. No credit is assumed for VP containment isolation in
the FHA within containment. The Containment Purge (VP) System and
Hydrogen Purge (VY) System containment isolation valves are sealed
closed during modes 1 through 4. The Containment Air Release and
Addition (VQ) System containment isolation valves are designed to
close within 5 seconds of a containment phase ``A'' isolation
signal. The prevention and mitigation of these accidents is not
affected by this change.
Test data demonstrates that the likelihood of a malfunction of a
resilient seal in one of the VP, VY, or VQ valves is not increased
by this change in the surveillances. The systems will continue to be
able to perform their design functions of isolating containment
during the evaluated accidents. Test procedures will continue to
monitor the leakage of these valves to ensure the design function
will continue to be met. There is no impact on previously evaluated
accidents since the valves will continue to close and seal or remain
closed as originally assumed in the accident scenarios.
Therefore, the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Second Standard
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No.
This change does not involve a physical alteration to the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing any normal plant operation. The
change does not alter assumptions made in the safety analyses or
licensing basis. This change will not affect or degrade the ability
of the Containment Purge System, Hydrogen Purge System, or
Containment Air Release and Addition System valves to perform their
specified safety functions. Therefore, the change does not create
the possibility of a new or different kind of credible accident from
any accident previously evaluated.
Third Standard
Does the proposed change involve a significant reduction in a
margin of safety?
No.
SR 3.6.3.6 currently states: ``The measured leakage rate for
Containment Purge System and Hydrogen Purge System valves must be <
0.05 La (Design Leakage Rate) when pressurized to Pa
(Design Containment Pressure). The measured leakage rate for
Containment Air Release and Addition valves must be < 0.01
La when pressurized to Pa. These required
maximum leak rates will not be changed by this amendment. Testing of
these valves to measure leakage through the valve seats will
continue, only at a different frequency based on past test results.
This will be a nominal frequency of 18 months for the VP System and
in accordance with 10 CFR 50, Appendix J, Option B for the VQ and VY
Systems. Therefore, the proposed changes listed above do not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 76488]]
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: July 19, 2004.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.8.4, ``DC Sources--Operating'' and TS
3.8.6, ``Battery Cell Parameters'' to allow for the replacement of the
existing nickel cadmium diesel generator batteries with conventional
lead acid batteries.
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 license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The DG batteries are not accident initiating equipment; they are
accident mitigating equipment. As such, they cannot affect the
probability of any accident being initiated. The performance of the
replacement batteries will exceed that of the existing batteries.
Therefore, no accident consequences will be adversely impacted.
(2) The proposed license amendments do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The DG batteries are not capable by themselves of initiating any
accident. Other than the replacement of the batteries themselves and
the associated modification work (e.g., installation of the battery
HVAC system), no physical changes to the overall plant are being
proposed. No changes to the overall manner in which the plant is
operated are being proposed. Therefore, no potential for new
accident types is generated.
(3) The proposed license amendments do not involve a significant
reduction in a margin of safety.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant
system, and the containment. The modification to replace the DG
batteries will not have any impact on these barriers. In addition,
no accident mitigating equipment will be adversely impacted as a
result of the battery replacement. The replacement batteries will
have overall performance capabilities equal to or greater than those
for the existing batteries. Therefore, existing safety margins will
be preserved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: September 21, 2004.
Description of amendments request: The proposed amendment would
delete the requirements from the technical specifications (TS) to
maintain hydrogen recombiners and hydrogen and oxygen monitors.
Licensees were generally required to implement upgrades as described in
NUREG-0737, ``Clarification of TMI [Three Mile Island] Action Plan
Requirements,'' and Regulatory Guide (RG) 1.97, ``Instrumentation for
Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs
Conditions During and Following an Accident.'' Implementation of these
upgrades was an outcome of the lessons learned from the accident that
occurred at TMI Unit 2. Requirements related to combustible gas control
were imposed by Order for many facilities and were added to or included
in the TS for nuclear power reactors currently licensed to operate. The
revised 10 CFR 50.44, ``Combustible gas control for nuclear power
reactors,'' eliminated the requirements for hydrogen recombiners and
relaxed safety classifications and licensee commitments to certain
design and qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration determination for referencing in
license amendment applications in the Federal Register on September 25,
2003 (68 FR 55416). The licensee affirmed the applicability of the
model no significant hazards consideration determination in its
application dated September 21, 2004.
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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG 1.97, Category 1, is intended for key variables that
most directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen and oxygen monitors no
longer meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44, the Commission found that
Category 3, as defined in RG 1.97, is an appropriate categorization
for the hydrogen monitors because the monitors are required to
diagnose the course of beyond design-basis accidents. Also, as part
of the rulemaking to revise 10 CFR 50.44, the Commission found that
Category 2, as defined in RG 1.97, is an appropriate categorization
for the oxygen monitors, because the monitors are required to verify
the status of the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
classification of the oxygen monitors as Category 2 and removal of
the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the SAMGs [severe
accident management guidelines], the emergency plan (EP), the
emergency operating procedures (EOPs), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TS, does
not involve a
[[Page 76489]]
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen monitor equipment are not considered accident precursors,
nor does their existence or elimination have any adverse impact on
the pre-accident state of the reactor core or post accident
confinement of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI Unit 2, accident can
be adequately met without reliance on safety-related hydrogen
monitors. Category 2 oxygen monitors are adequate to verify the
status of an inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI Unit 2, accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: April 14, 2004.
Description of amendment request: The proposed amendment would add
a new section to the Technical Specifications (TSs) and two new
Limiting Conditions for Operations (LCOs) to allow certain reactor
coolant system (RCS) hydrostatic and system leakage pressure tests to
be performed with the reactor pressure vessel temperature above
212[deg] Fahrenheit (F). The first LCO would allow specified TS
requirements to be changed to permit performance of special tests and
operations, which otherwise could not be performed if required to
comply with the requirements of the TSs. The second LCO would require
reactor low water level instrumentation, standby gas treatment system,
and secondary containment to be OPERABLE to allow certain RCS pressure
tests to be performed with the reactor pressure vessel temperature
above 212[deg] F, and provides for an exemption from the requirements
for OPERABILITY for other systems that currently go into effect when in
Hot Shutdown or when RCS temperature is greater than 212[deg] F. It
will also update the Table of Contents to reflect the proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The Nuclear Regulatory Commission (NRC) staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
probability of an accident previously evaluated is not significantly
increased because the proposed change will not alter the method by
which RCS hydrostatic pressure and leak testing is performed. Under
this proposed change the secondary containment, standby gas treatment
system and associated initiation instrumentation are required to be
operable during the performance of RCS hydrostatic pressure and leak
testing and would be capable of handling any airborne radioactivity or
steam leaks that could occur. The required pressure testing conditions
provide adequate assurance that the consequences of a steam leak will
be conservatively bounded by the consequences of a main steamline break
(MSLB) outside the primary containment. Accordingly, the consequences
of previously evaluated accidents are not increased significantly.
The proposed update to the Table of Contents is editorial in
nature. Since this update is administrative in nature, it cannot
increase the probability or consequences of a previously analyzed
accident.
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 previously evaluated?
The proposed amendment change will not alter the way that
hydrostatic pressure and leak testing is performed. Therefore, the
proposed change will 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?
The proposed amendment will not involve a significant reduction in
a margin of safety for a postulated MSLB outside of primary
containment. The proposed changes and additions result in increased
system operability requirements above those that currently exist during
the performance of RCS hydrostatic pressure and leak testing. The
incremental increase in stored energy in the vessel during testing will
be conservatively bounded by the consequences of the postulated MSLB
outside of primary containment. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The proposed update to the Table of Contents is editorial in
nature. Since this update is administrative in nature, the proposed
change does not involve a significant reduction in a margin of safety
Based on this review, it appears that the three standards of 10 CFR
50.92(c)) are satisfied. Therefore, the NRC staff proposes to determine
that the
[[Page 76490]]
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: Darrell Roberts.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: September 2, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification 4.5.B.2.2 (TS) to change the
surveillance requirement frequency for air testing the drywell and
suppression pool (torus) spray headers and nozzles from ``once every 5
years'' to ``following maintenance that could result in nozzle
blockage.''
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 drywell and torus headers and spray nozzles are not assumed
to be initiators of any accidents previously evaluated. Maintenance
practices and normal environmental conditions to which the system is
subjected are adequate to ensure operability of the systems. Since
the system will be able to perform its accident mitigation function,
the consequences of accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident [from any accident] previously
[evaluated]?
Response: No.
The revised surveillance does not introduce any new mode of
plant operation, does not involve physical modification of the
plant, or any new operating modes, and cannot introduce new accident
initiators. 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.
Maintenance practices and normal environmental conditions to
which the system is subjected are adequate to ensure operability of
the systems. As the spray nozzles are expected to remain fully
capable of performing their post-accident mitigation function,
margin of safety is not reduced. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: Darrell J. Roberts.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: September 2, 2004.
Description of amendment request: The proposed amendment would
remove a license condition that currently requires the reactor not to
be operated for more than 24 hours if one recirculation loop is out of
service. It would revise Technical Specifications (TSs) to allow the
minimum critical power ratio (MCPR) safety limit to be changed for
single loop operations (SLOs). It would also revise the current jet
pump limiting condition for operation and surveillance requirements to
allow for the conduct of a TS required surveillance during SLOs. The
proposed amendment would modify the TSs to address SLO operating
conditions and restrictions, and delete a TS condition related to
thermal-hydraulic stability. It would update the TSs for average planar
linear heat generation rate for SLOs, and update the thermal power
applicability restrictions to be consistent with NUREG-1433, Revision
3, ``Standard Technical Specifications for General Electric Boiling-
Water Reactors.'' It would also revise the TSs for linear heat
generation rate and MCPR for thermal power applicability restrictions.
The proposed amendment makes an administrative change to have MCPR
recalculated when reactor power is equal to or greater than 25 percent.
Lastly, it would update the TSs' table of contents and TS pages to
administratively reflect all of these proposed changes.
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 license and technical specification changes will
allow the plant to be operated with one recirculation pump for
longer than 24 hours provided that appropriate limits are
instituted. Extended single recirculation loop operation has been
evaluated and methodologies have been established for determining
appropriate operating limits. Implementation of the single
recirculation loop operating limits ensures that system operation is
in conformance with the conditions established to minimize the
probability of accidents and the associated consequences. Required
completion times for implementing the system operating limits and
restoring out of specification limits minimize the probability that
an accident occurs when out of specification conditions exist while
allowing for deliberate operator action.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence 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 license and technical specification changes will
allow plant operation with a single recirculation loop for longer
than 24 hours. The proposed changes introduce an additional
recirculation system-operating mode, however, existing system
component operating equipment or operating characteristics will not
change. The Pilgrim Station Single Loop Analysis Report identifies
required operating limits that apply when the system will be
operated in the single loop operation mode. Implementation of these
operating limits will ensure that the system is operated in
accordance with design. Additionally, revised jet pump surveillance
ensures that loop specific surveillance is performed as required to
validate the bounding assumptions of existing accident analyses. As
such, no new failure mechanisms are created and existing design
evaluations bound system operation.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident.
3. Does the proposed change involve a significant reduction in
[a] margin of safety?
Response: No.
The proposed license and technical specification changes
identify the operating limits that apply to single recirculation
loop operation. These proposed recirculation system limits were
identified to ensure that system operation would be in conformance
to the conditions evaluated in applicable accident and transient
analyses. Implementation of the proposed limits for single
recirculation loop operation ensures that safety margins are
maintained. Required completion times for implementing the system
operating limits minimizes the
[[Page 76491]]
possibility that an accident occurs when out of specification
conditions exist.
Therefore, the proposed changes do not involve a significant
reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: Darrell Roberts.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc. (licensee), Docket No. 50-271, Vermont Yankee Nuclear Power
Station, Vernon, Vermont
Date of amendment request: September 16, 2003 as supplemented by
letter dated March 15, 2004.
Description of amendment request: The proposed amendment would
relocate the current definition of surveillance frequency to new
Technical Specification (TS) Sections 4.0.2 and 4.0.3, and revise the
requirements for missed surveillance in Section 4.0.3. This change is
consistent with NRC-approved Industry/Technical Specification Task
Force (TSTF) change TSTF-358, Revision 5. The proposed change would
allow a longer period of time to perform a missed surveillance. The
time is extended from the current limit of up to 24 hours or up to the
limit of the specified frequency, whichever is less; to up to 24 hours
or up to the limit of the specified frequency, whichever is greater. In
conjunction with the proposed change, the proposed amendment would add
the requirements for a Bases Control Program which is consistent with
Section 5.5 of NUREG 1433. In addition, the current definition of
surveillance interval (definition ``Z'') would be re-worded and
relocated to new Section 4.0.1 consistent with Surveillance Requirement
3.0.1 of NUREG 1433. Appropriate Bases, also consistent with NUREG 1433
would be adopted for the new sections. An editorial change would be
made to TS 6.7.C which references the current definition of
surveillance frequency to now reference the new Section 4.0.2.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, 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 September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of the model NSHC determination in its application dated
September 16, 2003. The model NSHC determination analysis for changes
to the TS associated with missed surveillances, and the NSHC
determination analysis provided by the licensee for the remaining TS
changes, is provided herein.
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
With regard to the proposed change to the TS associated with
missed surveillances, the proposed change relaxes the time allowed
to perform a missed surveillance. The time between surveillances is
not an initiator of any accident previously evaluated. Consequently,
the probability of an accident previously evaluated is not
significantly increased. The equipment being tested is still
required to be operable and capable of performing the accident
mitigation functions assumed in the accident analysis. As a result,
the consequences of any accident previously evaluated are not
significantly affected. Any reduction in confidence that a standby
system might fail to perform its safety function due to a missed
surveillance is small and would not, in the absence of other
unrelated failures, lead to an increase in consequences beyond those
estimated by existing analyses. The addition of a requirement to
assess and manage the risk introduced by the missed surveillance
will further minimize possible concerns. Therefore, this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
With regard to the remaining proposed changes to the TSs, the
proposed changes do not involve physical changes to the plant or
introduce any new modes of operation. Accordingly, continued
assurance is provided that the process variables, structures,
systems, and components are maintained such that there will be no
degradation of any fission product barrier which could increase the
radiological consequences of an accident. Therefore, the proposed
changes do 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
With regard to the proposed changes to the TSs associated with
missed surveillances, the proposed change does not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) or a change in the methods governing
normal plant operation. A missed surveillance will not, in and of
itself, introduce new failure modes or effects and any increased
chance that a standby system might fail to perform its safety
function due to a missed surveillance would not, in the absence of
other unrelated failures, lead to an accident beyond those
previously evaluated. The addition of a requirement to assess and
manage the risk introduced by the missed surveillance will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
With regard to the remaining proposed changes to the TSs, 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. Thus, the
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
With regard to the proposed changes to the TSs associated with
missed surveillances, the extended time allowed to perform a missed
surveillance does not result in a significant reduction in the
margin of safety. As supported by the historical data, the likely
outcome of any surveillance is verification that the LCO [limiting
condition for operation] is met. Failure to perform a surveillance
within the prescribed frequency does not cause equipment to become
inoperable. The only effect of the additional time allowed to
perform a missed surveillance on the margin of safety is the
extension of the time until inoperable equipment is discovered to be
inoperable by the missed surveillance. However, given the rare
occurrence of inoperable equipment, and the rare occurrence of a
missed surveillance, a missed surveillance on inoperable equipment
would be very unlikely. This must be balanced against the real risk
of manipulating the plant equipment or condition to perform the
missed surveillance. In addition, parallel trains and alternate
equipment are typically available to perform the safety function of
the equipment not tested. Thus, there is confidence that the
equipment can perform its assumed safety function. Therefore, these
changes do not involve a significant reduction in a margin of
safety.
With regard to the remaining proposed changes to the TSs, the
administrative changes do not alter the basic operation of process
variables, systems, or components as described in the safety
analysis. No new equipment is introduced. Accordingly, the
[[Page 76492]]
proposed changes do not involve a significant reduction in a margin
of safety.
Therefore, this change does not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and its
endorsement of the model NSHC for missed surveillances 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. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: Allen G. Howe.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: October 5, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification Section 6.7.C ``Primary Containment Leak
Rate Testing Program,'' to allow a one-time extension to the 10-year
interval for performing the next Type A containment integrated leak
rate test (ILRT). Specifically, the change would allow the test to be
performed within 15 years from the last ILRT, which was performed in
April 1995.
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 operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed revision to Technical Specifications adds a one-
time extension to the current interval for Type A testing. The
current test interval of 10.6 years, based on past performance, is
extended on a one-time basis to fifteen years from the last Type A
test. The proposed extension to Type A testing cannot increase the
probability of an accident previously evaluated since the
containment Type A testing extension is not a modification and the
test extension is not of a type that could lead to equipment failure
or accident initiation.
The proposed extension to Type A testing does not involve a
significant increase in the consequences of an accident since
research documented in NUREG-1493 has found that, generically, very
few potential containment leakage paths are not identified by Type B
and C tests. The NUREG concluded that reducing the Type A (ILRT)
testing frequency to once per twenty years was found to lead to an
imperceptible increase in risk. These generic conclusions were
confirmed by a plant specific risk analysis performed using the
current Vermont Yankee Probabilistic Safety Assessment (PSA)
internal events model that concluded the consequences are low to
negligible.
Testing and inspection programs in place also provide a high
degree of assurance that the containment will not degrade in a
manner detectable only by Type A testing. The last two successful
Type A tests indicate a very leak tight containment. Type B and C
testing required by Technical Specifications will identify any
containment opening such as valves that would otherwise be detected
by the Type A tests. Inspections, including those required by the
ASME [American Society of Mechanical Engineers] code and the
Maintenance Rule are performed in order to identify indications of
containment degradation that could affect that leak tightness.
Therefore, the proposed changes do not represent a significant
increase in the probability or consequences of an accident
previously analyzed.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed revision to Technical Specifications adds a one
time extension to the current interval for Type A testing. The
current test interval of 10.6 years, based on past performance,
would be extended on a one time basis to fifteen years from the last
Type A test. The proposed extension to Type A testing cannot create
the possibility of a new or different type of accident since there
are no physical changes being made to the plant and there are no
changes to the operation of the plant that could introduce a new
failure mode creating an accident or affecting the mitigation of an
accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The proposed revision to Technical Specifications adds a one
time extension to the current interval for Type A testing. The
current test interval of 10.6 years, based on past performance,
would be extended on a one time basis to fifteen years from the last
Type A test. The proposed extension to Type A testing will not
significantly reduce the margin of safety. The NUREG-1493 generic
study of the effects of extending containment leakage testing found
that a 20-year extension in Type A leakage testing resulted in an
imperceptible increase in risk to the public. NUREG-1493 found that,
generically, the design containment leakage rate contributes about
0.1 percent to the individual risk and that the decrease in Type A
testing frequency would have a minimal affect on this risk since 95%
of the potential leakage paths are detected by Type C testing. This
was further confirmed by a plant specific risk assessment using the
current Vermont Yankee PSA internal events model that concluded the
risk associated with this change is negligibly small and/or non-risk
significant.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
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[NOTICE][PREAMB][AGENCY]*[/AGENCY][SUBJECT]*[/SUBJECT] ?>
NRC Section Chief: Allen G. Howe.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: October 6, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement 4.5.B.1 related
to air testing of the drywell spray headers and nozzles. Specifically,
the amendment would change the test frequency from once every 5 years
to following maintenance that could result in nozzle blockage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has
[[Page 76493]]
reviewed the licensee's analysis against the three standards of 10 CFR
50.92(c). The NRC staff's analysis is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would revise the Technical Specification
surveillance requirements associated with the air test of the drywell
spray headers and nozzles. The frequency of the air test would be
changed from a fixed 5-year frequency to following maintenance that
could result in nozzle blockage.
This surveillance test is performed while the plant is in a cold
shutdown condition and the equipment is not required to be operable.
The testing is to verify that the spray headers and nozzles are not
obstructed. The proposed change in the surveillance test frequency will
not result in any design changes to systems, structures, or components,
or their method of operation. The drywell spray headers and nozzles are
not initiators of any accidents previously evaluated. Therefore, the
proposed change does not involve a significant increase in the
probability of any accident previously evaluated.
The drywell spray headers provide a means to control both
temperature and pressure inside the primary containment, within design
limits, under post-accident conditions. Due to the system design and
operation considerations discussed in the licensee's application, the
potential for corrosion product formation is minimized. In addition,
the Vermont Yankee foreign material exclusion program has been judged
to be sufficient to ensure that foreign material is not inadvertently
introduced into the system. The proposed testing requirements are
considered sufficient to provide a high degree of confidence that
containment spray will function when required. Therefore, the proposed
change does not involve a significant increase in the consequences of
any 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 in the surveillance test frequency does not
create the possibility of a new or different kind of accident, since
there are no physical changes being made to the plant and there are no
changes to the operation of the plant that could introduce a new
failure mode, creating an accident or affecting the mitigation of an
accident.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the surveillance requirement to verify
that the drywell spray headers and nozzles are unobstructed. Industry
experience, Vermont Yankee surveillance history and the environmental
conditions the system is subjected to are adequate to ensure continued
system availability. As the spray nozzles are expected to remain
unobstructed and be able to perform their post-accident function, plant
safety is not affected.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92 are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: Allen G. Howe.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: July 9, 2004.
Description of amendment request: The Humboldt Bay Power Plant
(HBPP), Unit 3, is a decommissioning nuclear power plant that was
permanently shutdown in July 1976. In December of 2003, Pacific Gas and
Electric (PG&E or the licensee) applied for a license to store its
spent fuel in an onsite dry cask independent spent fuel storage
installation (ISFSI). Moving the spent fuel to an ISFSI would permit
the licensee to begin significant decommissioning activities. The
licensee has chosen to use a Holtec HI-STAR HB spent fuel cask handling
system involving a spent fuel multipurpose canister and overpack. To
facilitate spent fuel transfer from the HBPP spent fuel pool to the
ISFSI, the licensee will also need to install a new crane that can be
used to lift the cask handling system loaded with spent fuel
assemblies. The licensee states it will be able to satisfy the
applicable guidance of NUREG-0612, ``Control of Heavy Loads at Nuclear
Power Plants,'' and NUREG-0554. ``Single-Failure Proof Cranes for
Nuclear Power Plants,'' in performing the necessary movement of the
HBPP spent fuel to dry cask storage. The licensee has requested a
license amendment that approves the use of the crane and associated
changes to the HBPP Defueled Safety Analysis Report (DSAR) along with
analyses, design, and procedural changes required to implement transfer
of the spent fuel from the spent fuel pool to the ISFSI.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. With the HI-STAR HB System and the associated design and
handling procedures, all cask drops and other events, which could
damage other spent fuel, have been precluded through the robust
handling systems, and mechanical arrangement that preclude crane
movement over spent fuel, meeting the guidelines of NUREG-0612.
Revisions of the HBPP procedures implementing the control of heavy
loads ensures that PG&E will meet the NUREG-0612 guidelines and will
protect the fuel storage locations and the new HI-STAR HB System
loading/unloading activities. As a result of this design approach, a
cask-handling accident that results in a significant offsite
radiological release is not considered credible as demonstrated by
the probabilistic evaluation that was performed using the guidelines
of NUREG-0612 Appendix B and updated information from NUREG-1774
[``A Survey of Crane Operating Experience at U.S. Nuclear Power
Plants from 1968 through 2002.'']
Other HBPP licensing-basis events, such as the drop of a spent
fuel assembly, have not been affected by these changes and remain
bounding events for potential radiological consequences.
The proposed design of the dry cask system, the handling system,
and associated procedural controls provide assurance that: (1)
operational errors and mishandling events, and (2) support system
malfunctions will not result in an increase in the probability or
consequence of an accident previously analyzed.
The proposed changes to use the Holtec HI-STAR HB system have
been evaluated for seismic events and tornado missile impacts and it
has been determined that these changes will not result in an
increase in the probability or consequences of an accident
previously evaluated. The Fire Protection Program will ensure that
the combustible materials are properly controlled such that the
total combustibles meet the current program commitments. Therefore,
the proposed changes do not involve a
[[Page 76494]]
significant increase in the probability or consequences of an
accident.
2. Does the proposed amendment create the possibility of a new
or different type of accident from any accident previously
evaluated?
No. The engineering design measures and the handling procedures
preclude the possibility of new or different kinds of accidents.
Damage to 10 CFR 50 structures, systems, and components from the
cask handling and associated activities, and events resulting from
possible damage to contained fuel have been considered. Both the
types of accidents and the results remain within the envelope of
existing HBPP DSAR licensing basis analyses, as demonstrated by the
PG&E and Holtec analyses.
The rupture of multipurpose canister (MPC) dewatering, forced
helium dehydration or related closure system lines or the
malfunction of equipment during cask handling operations resulting
in radiological consequences are bounded by the HBPP DSAR fuel-
handling accident analysis.
Other design considerations, such as spent fuel pool (SFP)
thermal, water chemistry and clarity, criticality, and structural,
were evaluated and determined not to introduce the possibility of a
new or different kind of accident from any previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. With the Holtec HI-STAR HB System, and the associated design
and handling procedures, cask drops and other events have been
precluded through robust load handling systems, providing defense-
in-depth as described in NUREG-0612. Cask tipovers, while not
considered credible, are shown to be below the 60g limit, preventing
damage to the contained fuel assemblies (and associated structures),
and meeting the analysis guidelines of NUREG-0612. As the existing
licensing basis assumes a nonmechanistic drop damaging the SFP and
all fuel, the result of this design approach with the minimization
of drops and the associated structural challenges assure the margin
of safety has been maintained.
Other HBPP licensing-basis events, such as the drop of a spent
fuel assembly, have not been affected by these changes and remain
bounding events. Revision of HBPP procedures implementing the
control of heavy loads to incorporate the additional restrictions on
heavy loads movement will not affect the procedures or methodology
used and will, therefore, not affect margins.
Adverse effects from seismic events and/or cask drops or
tipovers have been evaluated, assuring that the fuel, MPC, and
overpack remain within their design bases. Since design basis
criteria are fully satisfied, there is no impact on the margin of
safety.
The Fire Protection Program will continue to ensure that the
combustible materials are properly controlled such that the total
combustibles meet the current program commitments. Thus, there are
no significant reductions in margin of safety associated with these
changes.
Other design considerations, such as SFP thermal, water
chemistry, criticality, and structural, were evaluated and
determined to not involve a reduction in a margin of safety.
Based on the above evaluations, the licensee concludes that the
activities associated with the above changes present no significant
hazards consideration under the standards set forth in 10 CFR 50.92 and
accordingly, a finding by the NRC of no significant hazards
consideration is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: October 6, 2004
Brief description of amendments: The proposed change will revise
the Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to
allow surveillance testing of the onsite diesel generators (DGs) during
power operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The licensee's analysis 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 design of plant equipment is not being modified by the
proposed changes. In addition, the DGs and their associated
emergency loads are accident mitigating features. As such, testing
of the diesel generators (DGs) themselves is not associated with any
potential accident-initiating mechanism. Therefore, there will be no
significant impact on any accident probabilities by the approval of
the requested changes.
The changes include an increase in the online time that a DG
under test will be paralleled to the grid (for SRs [Surveillance
Requirements] 3.8.1.10 and 3.8.1.14) or unavailable due to testing
(per SR 3.8.1.13). However, the overall time that the DG is
paralleled in all modes (outage/non-outage) should remain unchanged.
As such, the ability of the tested DG to respond to a design basis
accident [DBA] could be adversely impacted by the proposed changes.
However, the impacts are not considered significant based, in part,
on the ability of the remaining DG to mitigate a DBA or provide safe
shutdown. With regard to SR 3.8.1.10 and SR 3.8.1.14, experience
shows that testing per these SRs typically does not perturb the
electrical distribution system and share the same electrical
configuration alignment as the current monthly surveillance. In
addition, operating experience and qualitative evaluation of the
probability of the DG or bus loads being adversely affected
concurrent with or due to a significant grid disturbance, while the
DG is being tested, support the conclusion that the proposed changes
do not involve any significant increase in the likelihood of a
safety-related bus blackout or damage to plant loads.
The SR changes that are consistent with TSTF [Technical
Specification Task Force]-283 have been approved generically and for
individual Licensees. The on-line tests allowed by the TSTF are only
to be performed for the purpose of establishing OPERABILITY.
Performance of these SRs during restricted MODES will require an
assessment to assure plant safety is maintained or enhanced.
Deletion of expired TS LCO [Limiting Condition for Operation]
3.8.1, Required Action A.3, one-time 21-day Completion Time
allowance for Startup Transformer XST2 preventive maintenance is an
administrative change only. 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 changes would not create any new accidents since no
changes are being made to the plant that would introduce any new
accident causal mechanisms. Equipment will be operated in the same
configuration as currently allowed for other DG SRs that allow
testing during at-power operation. Deletion of expired TS LCO 3.8.1,
Required Action A.3, one-time 21-day Completion Time allowance for
Startup Transformer XST2 preventive maintenance is an administrative
change only. This license amendment request does not impact any
plant systems that are accident initiators; neither does it
adversely impact any accident mitigating systems.
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.
The proposed changes do not involve a significant reduction in
the margin of safety. The margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The proposed changes do not
directly affect these barriers, nor do
[[Page 76495]]
they involve any significant adverse impact on the DGs which serve
to support these barriers in the event of an accident concurrent
with a loss of offsite power. The proposed changes to the testing
requirements for the plant DGs do not affect the OPERABILITY
requirements for the DGs, as verification of such OPERABILITY will
continue to be performed as required (except during different
allowed MODES). The changes have an insignificant impact on DG
availability, as continued verification of OPERABILITY supports the
capability of the DGs to perform their required function of
providing emergency power to plant equipment that supports or
constitutes the fission product barriers. Only one DG is to be
tested at a time, so that the remaining DG will be available to
safely shut down the plant if required. Consequently, performance of
the fission product barriers will not be impacted by implementation
of the proposed amendment.
In addition, the proposed changes involve no changes to
setpoints or limits established or assumed by the accident analysis.
On this and the above basis, no safety margins will be impacted.
Deletion of expired TS LCO 3.8.1, Required Action A.3, one-time
21-day Completion Time allowance for Startup Transformer XST2
preventive maintenance is an administrative change only.
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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Mike Webb, Acting Chief.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: October 13, 2004.
Brief description of amendments: The proposed changes will revise
the Technical Specifications (TSs) to incorporate two topical reports
used to determine the core operating limits of Comanche Peak Steam
Electric Station (CPSES), Units 1 and 2, and delete reference to four
topical reports and a reference to NUREG-0800 that are no longer
required to support CPSES, Units 1 and 2, core operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The licensee's analysis 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 change is administrative in nature and as such does
not impact the condition or performance of any plant structure,
system or component. The core operating limits are established to
support Technical Specifications 3.1, 3.2, 3.3, and 3.4. The core
operating limits ensure that fuel design limits are not exceeded
during any conditions of normal operation or in the event of any
Anticipated Operational Occurrence (AOO). The methods used to
determine the core operating limits for each operating cycle are
based on methods previously found acceptable by the NRC and listed
in TS section 5.6.5.b. Application of these approved methods will
continue to ensure that acceptable operating limits are established
to protect the fuel cladding integrity during normal operation and
AOOs. The requested Technical Specification changes do not involve
any plant modifications or operational changes that could affect
system reliability, performance, or possibility of operator error.
The requested changes do not affect any postulated accident
precursors, do not affect any accident mitigation systems, and do
not introduce any new accident initiation mechanisms.
As a result, the proposed change to the CPSES Technical
Specifications does not involve any increase in the probability or
the consequences of any accident or malfunction of equipment
important to safety previously evaluated since neither accident
probabilities nor consequences are being affected by this proposed
administrative change.
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 is administrative in nature, and therefore
does not involve any changes in station operation or physical
modifications to the plant. In addition, no changes are being made
in the methods used to respond to plant transients that have been
previously analyzed. No changes are being made to plant parameters
within which the plant is normally operated or in the setpoints,
which initiate protective or mitigative actions, and no new failure
modes are being introduced.
Therefore, the proposed administrative change to the CPSES
Technical Specifications does not create the possibility of a new or
different kind of accident or malfunction of equipment important to
safety from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is administrative in nature and does not
impact station operation or any plant structure, system or component
that is relied upon for accident mitigation. Furthermore, the margin
of safety assumed in the plant safety analysis is not affected in
any way by the proposed administrative change.
Therefore, the proposed change to the CPSES Technical
Specifications does not involve any 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Michael Webb, Acting Chief.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: November 8, 2004.
Description of amendment request: To revise Technical Specification
Section 4.4.5.4 to modify the definitions of steam generator (SG) tube
``Plugging Limit'' and ``Tube Inspection.''
Date of publication of individual notice in the Federal Register:
November 24, 2004 (69 FR 68408).
Expiration date of individual notice: January 24, 2005.
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.
[[Page 76496]]
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].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: April 1, 2004.
Brief description of amendment: The amendment revises Technical
Specification (TS) requirements to adopt the provisions of the TS Task
Force (TSTF) change TSTF-359, regarding increased flexibility in mode
changes. The availability of TSTF-359 for adoption by licensees was
announced in the Federal Register on April 4, 2003 (68 FR 16579).
Date of issuance: November 29, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 163.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 24, 2004 (69 FR
52037).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 29, 2004.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: October 7, 2004, as supplemented by
letters dated November 12 and 18, 2004.
Description of amendment request: The amendment revised the Safety
Limit Minimum Critical Power Ratio in Technical Specification 2.1.1.2
to reflect the results of cycle-specific calculations performed for
Fermi 2 operating Cycles 10 and 11.
Date of issuance: November 30, 2004.
Effective date: As of the date of issuance and shall be implemented
prior to startup for Fermi 2 Cycle 11 operation.
Amendment No.: 164.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. (November 9, 2005; 69 FR 64986) The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by January 10, 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, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated November 30, 2004.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
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: January 26, 2004, as
supplemented September 13, 2004.
Brief description of amendments: These amendments authorized
changes to the BVPS-1 and 2 Updated Final Safety Analysis Reports
(UFSARs) to revise the level of the Ohio River that is assumed at the
onset of an accident during power operation to be 654.0' mean sea level
(msl) instead of 649.0' msl for BVPS-1 and 2. The proposed change is
consistent with current Technical Specification 3.7.5.1, which requires
the plant to shut down when the Ohio River reaches a level below 654.0'
msl.
Date of issuance: November 29, 2004.
Effective date: As of the date of issuance and shall submit the
changes authorized by these amendments with the next update of the
UFSARs in accordance with 10 CFR 50.71(e).
Amendment Nos.: 264 and 145.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
authorize changes to the UFSARs.
Date of initial notice in Federal Register: March 16, 2004 (69 FR
12369).
The supplement dated September 13, 2004, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 29, 2004 .
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: November 21, 2003, as
supplemented by letters dated May 18, and August 23, 2004.
Brief description of amendments: These amendments revised the St.
Lucie Unit 1 and 2 Technical Specifications (TSs) to eliminate certain
pressure sensor response time testing requirements. Elimination of
these tests is discussed in the Combustion Engineering Owners Group
Topical Report CE NPSD-1167, Revision 2, ``Elimination of Pressure
Sensor
[[Page 76497]]
Response Time Testing Requirements,'' which was approved by the NRC
staff in letters dated July 24, 2000, and December 5, 2000.
Specifically, these amendments revise the St. Lucie Units 1 and 2 TS
Definitions 1.12, ``Engineered Safety Features Response Time,'' and
1.26, ``Reactor Protection System Response Time.''
Date of Issuance: November 30, 2004.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 195 and 137
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the TSs.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57675).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 30, 2004.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: November 26, 2002, as
supplemented by letters dated September 8, 2003, October 30, 2003, June
21, 2004, and October 8, 2004.
Brief description of amendments: The amendments increased the total
spent fuel wet storage capacity for each unit, by adding a spent fuel
storage rack in the cask area in each unit's spent fuel pool. Each rack
increased both units' storage capacity by 131 fuel assemblies. The
amendments also included the addition of the design of the racks in
Section 5.6.1.1.c of the Technical Specifications (TSs), and revised
the stated spent fuel capacity in TS Section 5.6.3 and the location
called out in the Design Features Sections 5.6.1.1a and b of the TSs
referring to Updated Final Safety Analysis Report Appendix 14D rather
the Westinghouse Report WCAP-14416-P.
Date of issuance: November 24, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos: 226 and 222.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TSs.
Date of initial notice in Federal Register: January 28, 2003 (69 FR
4246). The supplemental letters provided clarifying information that
did not expand the scope of the original application or change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in an Environmental Assessment dated October 17, 2003, and in a Safety
Evaluation dated November 24, 2004.
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: August 17, 2004.
Brief description of amendment: The amendment revised Section
3.3.1, ``Oxygen Concentration,'' of the Technical Specifications to add
a new action, allowing 24 hours to restore the oxygen concentration
within the limit of <4% by volume if the limit is exceeded when the
reactor is operating in the power operating condition.
Date of Issuance: November 29, 2004.
Effective date: November 29, 2004 and shall be implemented within
15 days of issuance.
Amendment No.: 185.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53110).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated November 29, 2004.
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 12, 2003, and its
supplements dated April 23, June 4, and August 30, 2004.
Brief description of amendments: The amendments increase the
current steam generator narrow range water level-low low setpoints from
greater or equal to 7.0 percent allowable value and 7.2 percent nominal
trip setpoint to greater than or equal to 14.8 percent allowable value
and 15.0 percent nominal trip setpoint. The reactor trip setpoint is
specified in TS Table 3.3.1-1, ``Reactor Trip System Instrumentation,''
and the actuation setpoint to start the auxiliary feedwater pumps is
specified in TS Table 3.3.2-1, ``Engineered Safety Feature Actuation
System Instrumentation.''
Date of issuance: December 2, 2004.
Effective date: December 2, 2004, and shall be implemented within
90 days from the date of issuance.
Amendment Nos.: Unit 1-178; Unit 2-180.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 25, 2003 (68
FR 66138) The April 23, June 4, and August 30, 2004, 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 December 2, 2004.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: December 5, 2003, as
supplemented by letter dated June 4, 2004.
Brief description of amendments: The amendments revised SSES 1 and
2 Technical Specifications (TSs) by adding a requirement to apply
linear heat generation (LHGR) limits if the main turbine bypass system
becomes inoperable. The proposed changes clarify TS 3.7.6 to state that
both minimum critical power ratio and LHGR limits for an inoperable
main turbine bypass system are required if the system becomes
inoperable.
Date of issuance: December 3, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 218 and 193.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 6, 2004 (69 FR
698). The supplement dated June 6, 2004, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 3, 2004.
No significant hazards consideration comments received: No.
[[Page 76498]]
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: August 12, 2004, as superseded by letter
dated October 5, 2004, as supplemented by letter dated October 11,
2004.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3/4.4.5, in conjunction with the new administrative
control TS 6.8.3.o and reporting requirement TS 6.9.1.7, to establish a
new programmatic, largely performance-based framework for ensuring SG
tube integrity. The reactor coolant system leakage requirements of TS
3.4.6.2 are also revised.
Date of issuance: November 24, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 1--164; Unit 2--154.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53113). The October 5, 2004, letter which superseded the August 12,
2004, letter and the supplement dated October 11, 2004, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not
significantly 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 November 24, 2004.
No significant hazards consideration comments received: No.
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: August 26, 2004.
Brief description of amendments: The amendments eliminate the
requirements in the TS associated with hydrogen recombiners and
hydrogen monitors. A notice of availability for this TS improvement
using the consolidated line item improvement process was published in
the Federal Register on September 25, 2003 (68 FR 55416).
Date of issuance: November 30, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 1--165; Unit 2--155.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57996).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 30, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 13th day of December 2004.
For the Nuclear Regulatory Commission.
James E. Lyons,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 04-27614 Filed 12-20-04; 8:45 am]
BILLING CODE 7590-01-P