[Federal Register Volume 64, Number 144 (Wednesday, July 28, 1999)]
[Notices]
[Pages 40903-40914]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-19133]
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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 Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Public Law 97-415 revised section 189 of the Atomic Energy Act
of 1954, as amended (the Act), to require the Commission to publish
notice of any amendments issued, or proposed to be issued, under a new
provision of section 189 of the Act. This provision grants the
Commission the authority to issue and make immediately effective any
[[Page 40904]]
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 July 3, 1999, through July 16, 1999. The
last biweekly notice was published on July 14, 1999 (64 FR 38022).
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing 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 Administration 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 NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By August 27, 1999, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating 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 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition, and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which much include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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.
[[Page 40905]]
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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Stream
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: March 26, 1999.
Description of amendment request: The proposed change provides a
Required Action and Completion Time for the Ultimate Heat Sink (UHS) in
the event that service water temperature exceeds the current 95 deg.F
surveillance limit. It involves an allowance to continue operation for
a period of 8 hours with the UHS at a temperature greater than the
temperature limits provided in Technical Specification (TS) Limiting
Condition of Operation 3.7.8, ``Ultimate Heat Sink (UHS)'' and provides
an upper UHS temperature limit beyond which plant shutdown is required.
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:
Carolina Power & Light (CP&L) Company has evaluated the proposed
Technical Specification change and has concluded that it does not
involve a significant hazards consideration. The conclusion is in
accordance with the criteria set forth in 10 CFR 50.92. The bases
for the conclusion that the proposed change does not involve a
significant hazards consideration are discussed below.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components. The proposed change will
allow plant operation for a short period of time when the service
water temperature exceeds 95 deg.F. If the service water temperature
is restored within the allowed time, a plant shutdown is not
required. This minimizes plant transients, which reduces the
probability of a reactor trip and the resulting challenges to
mitigating systems. A service water temperature of up to 99 deg.F
does not increase the failure rate of systems, structures or
components because the systems, structures, and components are
designed for higher temperatures than at which they operate.
The Service Water (SW) System temperature is not assumed to be
an initiating condition of any accident evaluated in the safety
analysis report. Therefore, the allowance of a limited time for
service water temperature to be in excess of 95 deg.F does not
involve an increase in the probability of an accident previously
evaluated in the safety analysis report (SAR). The SW System
supports operability of safety related systems used to mitigate the
consequences of an accident. The service water temperature is not
expected to increase significantly beyond 95 deg.F due to the
limited time allowed by the proposed change in conjunction with the
generally slow rate of temperature increase experienced from thermal
changes in Lake Robinson. The capability of components to perform
their safety related function is not affected up to a service water
temperature of 99 deg.F with the exception of the Containment Air
Recirculation Fan Coolers. The heat removal capacity of the
Containment Air Recirculation Fan Coolers is not expected to be
significantly reduced by a small increase in service water
temperature. If heat removal is not significantly reduced,
containment pressure and leakage will not be significantly
increased, and the doses from containment leakage will not be
significantly increased. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated in the SAR.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components. A service water temperature
of up to 99 deg.F does not introduce new failure mechanisms of
systems, structures or components not already considered in the SAR
because the systems, structures, and components are designed for
higher temperatures than at which they operate. Therefore, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change will allow a small increase in service water
temperature above the design basis limit for the SW System and delay
by 8 hours the requirement to shutdown the plant when the service
water system design limit is exceeded. There are design margins
associated with systems, structures and components that are cooled
by the service water system that are affected. The capability of
components to perform their safety related function is not affected
up to a service water temperature 99 deg.F with the exception of the
Containment Air Recirculation Fan Coolers. The Containment Air
Recirculation Fan Coolers remove heat from containment to mitigate
containment pressure and temperature following a MSLB (main
streamline break) inside containment or a Large Break LOCA (loss-of-
coolant accident) inside containment. An increase in service water
temperature in excess of the design limit due to hot weather
conditions is expected to be small due to the limited time allowed
by the proposed change in conjunction with the generally slow rate
of temperature increase experienced from thermal changes in Lake
Robinson. Therefore, the effect on the Containment Air Recirculation
Fan Coolers' heat removal capacity and the resulting containment
pressure and temperature is expected to be small. Therefore, there
is no significant reduction in margin of safety associated with this
change.
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.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Sheri R. Peterson.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: June 29, 1999.
Description of amendment request: This amendment request proposes
to increase the notch testing surveillance interval of partially
withdrawn control rods in Technical Specification Surveillance
Requirement 3/4.3.C,
[[Page 40906]]
``Reactivity Control--Control Rod Operability,'' from an interval of
once in 7 days to once in 31 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
The proposed change extends the Surveillance Frequency for
partially withdrawn control rods. The change does not affect
equipment design or operation. The affected Surveillance is not
considered to be an accident initiator. Therefore, this change will
not significantly increase the probability of an accident previously
evaluated. Furthermore, extension of the Surveillance Frequency will
not impact the ability to perform its function following an
accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The extension of the Surveillance Frequency does not involve
physical modification to the plant and does not introduce a new mode
of operation.
Therefore, the change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
The change in the Surveillance Frequency only provides a minor
reduction in the probability of finding an inoperable control rod.
Most of the control rods will continue to be tested on the current
Frequency. However, if one stuck rod is identified, all rods must be
checked promptly.
Therefore, these changes do not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposed to determine that
the requested amendments involve no significant hazards
consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: April 1, 1999.
Description of amendment request: The proposed license amendment
would modify the Technical Specifications (TSs) to incorporate certain
improvements from the Revised Standard Technical Specifications for B&W
Plants (NUREG-1430) that would add limiting conditions for operation
action statements, make surveillance requirements more consistent with
the revised standard TSs, correct conflicts or inconsistencies from
earlier TS revisions, correct administrative errors, and revise the
spent fuel pool sampling from monthly and after adding chemicals to
weekly.
The staff's proposed no significant hazards determination below
does not address the licensee's proposed changes with respect to a high
pressure injection system operation in a low temperature overpressure
environment.
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. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated. The proposed amendment makes administrative corrections,
adds conditions to the limiting conditions of operation [LCOs],
revises selected time clocks and surveillance requirements
consistent with NUREG 1430, and adds a time clock to a unique LCO.
These changes have no effect on the plant design or operation. The
reliability of systems and components relied upon to prevent or
mitigate the consequences of accidents previously evaluated is not
degraded by proposed changes. Therefore, operation in accordance
with the proposed amendment does not involve a significant increase
in the probability of occurrence or consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated, because no new
accident initiators would be created.
3. Operation of the facility in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety because no changes to plant operating limits or limiting
safety system settings are proposed.
The NRC staff has reviewed the licensee's analysis and based on the
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.
Local Public Document Room Location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 22037.
NRC Section Chief: S. Singh Bajwa.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 4, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 Technical Specifications (TSS) proposes to revise
the definition of operating personnel in section 6.2.2.g to make it
consistent with the Standard Technical Specifications and to remove a
footnote.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licenses has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No, these TS changes are administrative in nature. Removing the
statement in section 6.2.2.g that defines on shift operating
personnel and adding a new paragraph consistent with the Standard
Technical Specifications is an administrative line item change that
follows NRC guidance. The current statement is not needed because TS
Table 6.2.1 defines the minimum operations shift crew composition
and commitments to Table B-1 of NUREG-0654 defines the minimum
staffing requirements for each function area.
The change to TS 6.2.2.i is administrative in nature. The
statement that reads, ``For the period ending three years after
restart from the 1993/1994 Performance Improvement Outage, the
Operations Manager will be permitted to have held a SRO [senior
reactor operator] license at a Pressurized Water Reactor other than
Indian Point Unit 3'', was a relaxation of the requirements of
6.2.2i.
Therefore, these changes will not increase the probability or
consequences of an accident previously evaluated, because they are
administrative and affect neither accident initiation or mitigation.
2. Does the proposed license amendment create the possibility of
a new or different
[[Page 40907]]
kind of accident from any accident previously evaluated?
No, these TS changes are administrative in nature. Removing the
statement in section 6.2.2.g that defines on shift operating
personnel and adding a new paragraph consistent with the Standard
Technical Specifications is an administrative line item change that
follows NRC guidance. The current statement is not needed because TS
Table 6.2-1 defines the minimum operations shift crew composition
and commitments to Table B-1 of NUREG-0654 defines the minimum
staffing requirements for each function area.
The change to TS 6.2.2.i is administrative in nature. The
statement that reads, ``For the period ending three years after
restart from the 1993/1994 Performance Improvement Outage, the
Operations Manager will be permitted to have held a SRO license at a
Pressurized Water Reactor other than Indian Point Unit 3'', was a
relaxation of the requirements of 6.2.2.i.
These changes are administrative, and do not affect how the
plant is operated. They also follow the guidance of the Standard
Technical Specifications. Therefore, these changes will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No, these TS change is administrative in nature. Removing the
statement in section 6.2.2.g that defines on shift operating
personnel and adding a new paragraph consistent with the Standard
Technical Specification is an administrative line item change that
follows NRC guidance. The current statement is not needed because TS
Table 6.2-1 defines the minimum operations shift new composition and
commitments to Table B-1 of NUREG-0654 defines the minimum staffing
requirements for each function area.
The change to TS 6.2.2.i is administrative in nature. The
statement that reads, ``For the period ending three years after
restart from the 1993/1994 Performance Improvement Outage, the
Operations Manager will be permitted to have held a SRO license at a
Pressurized Water Reactor other than Indian Point Unit 3'', was a
relaxation of the requirements of 6.2.2.i.
These changes are administrative, and do not affect how the
plant is operated. They also follow the guidance of the Standard
Technical Specifications. Therefore, these 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 50.92(c) are
satisfied. Therefore, the NRC staff proposed to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for Licensee; Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Description of amendment requests: The licensee proposed changes to
Technical Specification (TS) 3.3.5 ``ESFAS Instrumentation'' to include
restrictions on operation with a channel of the refueling water storage
tank level-low input to the recirculation actuation signal (RAS) and
the steam generator pressure-low input or steam generator pressure
difference-high input to the emergency feedwater actuation signal
(EFAS) in the tripped condition. The current TS allows plant operation
in this condition indefinitely. The licensee has determined that
unacceptable consequences could result from a spurious trip of RAS or
EFAS due to operation with a channel in trip condition. The licensee
states that the proposed TS changes would improve plant operational
safety and, thereby, reduce plant risk.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
This change provides limits for operating with a channel of the
Refueling Water Storage Tank (RWST) Level-Low input in the
Recirculation Actuation Signal (RAS) or the Steam Generator (SG)
Pressure-Low or SG Pressure Difference (SGPD)-High input to the
Emergency Feedwater Actuation Signal (EFAS) in trip.
As a result of this change, the potential for an inadvertent
actuation of either of these two signals is reduced. The proposed
Completion Times are based on Probabilistic Risk Assessment (PRA)
considerations, and are conservative compared to the current
unlimited Completion Times.
The consequences of an inadvertent actuation of EFAS or RAS are
unaffected by this change.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
(2) Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
This proposed change provides additional time limits on
operation with a channel of the RWST Level-Low input to RAS or the
SG Pressure-Lower SGPD-High inputs to EFAS in trip. Operation in
this condition is currently allowed indefinitely. The proposed
restrictions reduce the possibility of an inadvertent actuation of
RAS or EFAS, and do not allow operation in any configuration not
currently allowed by the Technical Specifications (TSs).
Therefore, this proposed change will not create the possibility
of a new or different kind of accident from any accident that has
been previously evaluated.
(3) Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The proposed change provides additional time limits on operation
with a channel of the RWST Level-Low input to RAS or the SG
Pressure-Low or SGPD-High inputs to RAS or EFAS in trip. The
proposed limits are conservative compared to the current
requirements, where the time limit is unrestricted. The overall
impact of the change will be [an] increase in the margin of safety.
Therefore, there will be no significant reduction in a margin of
safety as a result of this change.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of application for amendments: June 7, 1999 (TS 99-09).
Brief description of amendments: The proposed amendment would
change the Sequoyah Unit 2 Technical Specification (TS) requirements by
adding a new temporary Figure 3.4-1a and temporary footnotes to TS
3.4.8, ``Specific Activity,'' Table 4.4-4, and to corresponding Bases
in order to raise the reactor coolant specific activity limit to 1.0
microcurie per milligram Dose Equivalent iodine-131 for the remainder
of Unit 2 Cycle 10 operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority, the licensee, has provided its analysis of the issue of no
significant hazards
[[Page 40908]]
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed TS change increases the allowed reactor coolant
specific activity for iodine-131 and decreases the leakage quantity
that would be postulated to occur at the faulted steam generator
(SG) during a main steam line break (MSLB) accident. The described
changes will return these parameters to the same values under which
the plant operated prior to the implementation of TS Change 98-02
submitted on June 26, 1998. The June 26, 1998 submittal was a
voluntary change that allowed for a greater leakage quantity during
an MSLB accident as described in Generic Letter 95-05. Returning
these parameters to their previous values does not affect or
increase the probability of any accidents previously evaluated.
An increase in the consequences of an accident would not occur
because the proportional increase in reactor coolant specific
activity, while proportionally decreasing the allowable primary-to-
secondary leakage during a postulated MSLB accident to values under
which the plant was previously operated, was evaluated in [Topical
Report No.] WCAP-13990 during the establishment of the original
primary-to-secondary leak limits. No changes to the physical plant,
to the plant operation, or maintenance practices have been
implemented that would invalidate the limits defined in WCAP-13990.
The control room dose, the low population zone dose, and the
dose at the exclusion area boundary remain bounded by the acceptance
criteria of the Updated Final Safety Analysis Report. Therefore, the
proposed TS change does not result in an increase in the
consequences of an accident previously analyzed.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS change does not alter the configuration of the
plant. The changes do not directly affect plant operation. The
change will not result in the installation of any new equipment or
systems or the modification of any existing equipment or systems. No
new operating procedures, conditions, or modes will be created by
this proposed change. SG tube structural integrity, as defined in
draft Regulatory Guide 1.121, remains unchanged. Therefore, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
Raising the allowed reactor coolant specific activity, while
decreasing the allowed primary-to-secondary leakage during a
postulated MSLB accident, keeps the amount of activity released to
the environment unchanged. Design basis and offsite dose calculation
assumptions remain satisfied. Therefore, the proposed change does
not result in a significant reduction in the margin of safety.
The NRC 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.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: May 24, 1999, as supplemented by letter
dated July 9, 1999.
Brief description of amendments: The proposed license amendments
would remove several cycle-specific parameter limits from the Technical
Specifications (TSs). These parameter limits would be added to the Core
Operating Limits Report (COLR). Appropriate references to the COLR
would be inserted in the affected TSs. In addition, the core safety
limit curves would be replaced with safety limits more directly
applicable to the fuel and fuel cladding fission product barriers. The
affected Technical Specifications are: (1) TS 2.0, ``Safety Limits
(SLs),'' (2) TS 3.3.1, ``Reactor Trip System Instrumentation
Setpoints,'' (3) TS 3.4.1, ``RCS pressure temperature and flow from
Nucleate Boiling (DNB) Limits,'' and (4) TS 5.6.5, ``Core Operating
Limits Report.'' The May 24, 1999, application was previously noticed
and published in the Federal Register on June 30, 1999 (64 FR 53213).
The July 9, 1999, supplement provided proposed additional
information that would: (a) Add the Reactor Core Safety Limit figures
to the COLR, (b) clarify that the overpower N-16 setpoint remains in
the TSs, and (c) reflect NRC approval of the topical reports used to
determine the core operating limits presented in the COLR. The
supplemental information is being noticed herein to address the issue
of no significant hazards consideration.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes remove cycle-specific parameter limits from
the Technical Specifications, add them to the list of limits
contained in the Core Operating Limits Report (COLR), and revise the
Administrative Controls section of the Technical Specifications. The
proposed changes also insert the original minimum RCS flow limits
into the Technical Specifications. The changes do not, by
themselves, alter any of the parameter limits. The changes are
administrative in nature and have no adverse effect on the
probability of an accident or on the consequences of an accident
previously evaluated. The removal of parameter limits from the
Technical Specifications does not eliminate the requirement to
comply with the parameter limits.
The parameter limits in the COLR may be revised without prior
NRC approval. However, [Technical] Specification 5.6.5c continues to
ensure that the parameter limits are developed using NRC-approved
methodologies and that applicable limits of the safety analyses are
met. While future changes to the COLR parameter limits could result
in event consequences which are either slightly less or slightly
more severe than the consequences for the same event using the
present parameter limits, the differences would not be significant
and would be bounded by the requirement of specification 5.6.5c to
meet the applicable limits of the safety analysis.
Based on the above, addition of the minimum RCS flow limit into
the Technical Specifications, removal of the parameter limits the
Technical Specifications and the addition of the described limits in
the COLR, thus allowing revision of the parameter limits without
prior NRC approval, has no significant effect on the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes add the minimum RCS flow limit into the
Technical Specifications, remove certain parameter limits from the
Technical Specifications and add these limits to the list of limits
in the COLR, thus removing the requirements for prior NRC approval
of revisions to those parameters. The changes do not add new
hardware or change plant operations and therefore cannot initiate an
event nor cause an analyzed event to progress differently. Thus, the
possibility of a new or different kind of accident is not created.
3. Do the proposed changes involved a significant reduction in a
margin of safety?
The margin of safety is the difference between the acceptance
criteria and the associated failure values. The proposed changes do
not affect the failure values for any parameter. Though the accident
analyses, all applicable limits (i.e., relevant event acceptance
criteria as described in the NRC-approved analysis methodologies)
are shown to be satisfied; therefore, there is no impact
[[Page 40909]]
on event acceptance criteria. Because neither the failure values nor
the acceptance criteria are affected, the proposed change has no
effect on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont.
Date of amendment request: May 26, 1999.
Description of amendment request: The licensee proposed revising
the suppression pool water temperature surveillance requirements to
specify monitoring the temperature every 5 minutes when performing
testing that adds to the suppression pool. In addition, the licensee
proposed revising the requirement to check the suppression chamber
water level and temperature from ``once per shift'' to ``daily'' and
specify that it is the average temperature that is checked.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided the NCR its analysis of the issue of no significant hazard
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. 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.
Vermont Yankee has determined that the proposed change will not
involve a significant increase in the probability or consequences of an
accident previously evaluated. The proposed change revises the
surveillance frequency for ``once per shift'' suppression pool water
level and temperature monitoring. Additionally, the surveillance
requirement for suppression pool water temperature monitoring when
there are indications of relief valve operation that add heat to the
suppression pool is also revised. The proposed change will revise the
surveillance wording such that routine suppression pool monitoring will
be ``daily'' and an operator will verify pool temperature every 5
minutes only during testing that adds heat to the suppression pool.
Also clarified, is that the parameter being monitored is ``average''
suppression pool water temperature.
The consequence of an accident previously evaluated is not
significantly increased since the initial suppression pool water
temperature limit, which is an input valve for accident analyses, is
not changed.
The proposed change affects only surveillance requirements and does
not require any hardware or equipment modification. Equipment
operation, plant limiting conditions for operation, and accident
analyses will be unchanged. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
accidents.
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.
Vermont Yankee has determined that the proposed change does not
create the possibility of a new or different kind of accident from any
accident previously evaluated. The proposed change involves revision of
Technical Specification surveillance requirements. There are no
hardware modifications or equipment changes involved and operation of
plant equipment will be unchanged. Thus, no new or different accident
precursors will be created by this change.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not involve a significant
reduction in a margin or safety. VY has determined that the proposed
change does not involve a significant reduction in a margin of safety.
The proposed change involves revision of Technical Specification
surveillance requirements. There are no hardware modifications or
equipment changes involved and plant operation and accident analyses
are unchanged. The initial suppression pool water temperature limit,
which is an input value for accident analyses, is not changed.
Therefore, the proposed change will not involve a significant reduction
in the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attoney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: June 29, 1999.
Description of amendment request: The licensee proposed revising
the leak rate requirements of Technical Specifications 3.7.A.4 and
4.7.A.4 for the main steam line isolation valves. Specifically, a total
leakage rate allowable value for the sum of the four main steam lines
is proposed that is equal to four times the current individual main
steam line isolation valve leakage rate allowable value. The individual
main steam line isolation valve leakage rate allowable value is
proposed to be one half of the total leakage rate allowable value.
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 change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that contribute to the
initiation of any accidents previously evaluated. Thus, the proposed
change cannot increase the probability of any accident previously
evaluated.
The proposed change does not affect the leak-tight integrity of
the containment structure that is designed to mitigate the
consequences of a loss-of-coolant accident (LOCA). The primary
containment must maintain functional integrity during and following
the peak transient pressures and temperatures that result from any
LOCA, thereby limiting fission product leakage following the
accident. Because the proposed change does not alter any of the
fission product lead rate assumptions used in the design basis LOCA
analysis, the analyzed consequences of the Loss of Coolant Accident
are not changed.
The control room radiological habitability analysis uses as an
input assumption main steam line leakage rate at four times the
current Technical Specifications limit. An allowable value for total
main steam line
[[Page 40910]]
leakage rate equivalent to four times the current Technical
Specifications limit for a single main steam line isolation valve is
being added by this change. Thus, there is no effect on the main
control room radiological habitability calculation.
Based on the above VY [Vermont Yankee] has concluded that the
proposed change will not result in a significant increase in the
probability or consequences of any accident previously evaluated.
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 change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any parameters or conditions that could contribute to the
initiation of any accident. The methods of performing the tests are
not changed. No new accident modes are created. No safety-related
equipment or safety functions are altered as a result of this
change. Restating the acceptance criteria while maintaining the
assumptions of all affected calculations has no influence over nor
does it contribute to, the possibility of a new or different kind of
accident or malfunction from those previously evaluated.
Based on the above VY has concluded that the proposed change
will not create the possibility of a new or different kind of
accident from those previously evaluated.
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.
Restating the acceptance criteria for the main steam line
isolation valve leakage rate while maintaining the assumptions of
all affected calculations does not impact the margin of safety. The
0.6La maximum and minimum pathway leakage rate acceptance
criteria provide the previously analyzed margin of safety. The
testing method for determining the leak-tightness of the main steam
line isolation valves has not changed. The leak rate test results
are presently added to the Types B and C tests summation. The
0.6La maximum and minimum pathway leak rate acceptance
criteria and the proposed Technical Specifications requirements
provide assurance that component degradation does not impact the
assumptions used to determine, nor provide a reduction in, and the
analyzed margin of safety.
Based on the above VY has concluded that the proposed change
will not cause 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.
Local Public Document Room Location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trobridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont.
Date of amendment request: July 12, 1999.
Description of amendment request: The amendment would revise the
value for the Safety Limit Minimum Critical Power Ratio (SLMCPR) and
delete the wording specifying these as Cycle 20 values.
Basis for proposed no significant hazards consideration
determination: As required by to 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 basis of the SLMCPR is to ensure no mechanistic fuel damages
is calculated to occur if the limit is not violated. The new SLMCPR
values preserve the existing margin to transition boiling and
probability of fuel damage is not increased. The derivation of the
revised SLMCPR for Vermont Yankee for incorporation into the
Technical Specifications, and its use to determine plant and cycle-
specific thermal limits, have been performed using NRC approved
methods. These plant-specific calculations are performing each
operating cycle and if necessary, will require future changes to
these values based upon revised core designs. The revised SLMCPR
values do not change the method of operating the plant and have no
effect on the probability of an accident initiating event or
transient.
Based on the above, Vermont Yankee has concluded that the
proposed change will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
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 changes result only from a specific analysis for
the Vermont Yankee core reload design and deletion of a cycle
specific reference for the values. These changes do not involve any
new or different method for operating the facility and do not
involve any facility modifications. No new initiating events or
transients result from these changes.
Based on the above, Vermont Yankee has concluded that the
proposed change will not create the possibility of a new or
different kind of accident from those previously evaluated.
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 new SLMCPR is calculated using NRC approved methods with
plant and cycle specific parameters for the current core design. The
SLMCPR value remains high enough to ensure that greater than 99.9%
of all fuel rods in the core will avoid transition boiling if the
limit is not violated, thereby preserving the fuel cladding
integrity. The operating MCPR limit is set appropriately above the
safety limit value to ensure margin when the cycle specific
transients are evaluated.
As a result, Vermont Yankee has determined that the proposed
change will not result in 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.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: June 22, 1999 (TSCR 210).
Description of amendment request: The proposed amendments reflect
changes to the Point Beach Nuclear Plant (PBNP) Units 1 and 2 Technical
Specifications (TSs) in order to incorporate the Westinghouse 422V+
fuel assemblies into the PBNP reactor cores. 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. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant increase in
the probability or consequences of an accident previously evaluated.
The accidents which are potentially affected by the parameters
and assumptions associated with this amendment have been evaluated/
analyzed and all design standards and applicable safety criteria are
met. The consideration of these changes does not result in a
situation where the design and construction standards that were
applicable prior to the change are altered. Therefore, the changed
occurring with this amendment will not result in any additional
challenges to plant equipment that could increase the probability of
any previously evaluated accident.
The proposed changes associated with this amendment do not
affect plant systems such that their function in the control of
[[Page 40911]]
radiological consequences is adversely affected. The safety
evaluation (included in Attachment 2 of this submittal) documents
that the design standards and applicable safety criteria limits
continue to be met and therefore fission barrier integrity is not
challenged. The proposed changes have been shown not to adversely
affect the response of the plant to postulated accident scenarios.
Existing system and component redundancy and operation is not being
changed by these proposed changes. These changes will therefore not
affect the mitigation of the radiological consequences of any
accident described in the FSAR [final safety analysis report].
In some cases, the results of the revised radiological analyses
are greater than those of the current FSAR analysis. In other cases,
the new and old analyses are not directly comparable because the
radiological bases for the new analyses have been upgraded to meet
more current NRC requirements. However, in all cases, the calculated
doses are well within the regulatory acceptance criteria and do not
constitute an unacceptable significant increase in consequences.
Since the actual plant configuration, performance of systems, and
initiating event mechanisms are not being changed as a result of
this evaluation, the probability or consequences of an accident
previously evaluated is not significantly increased.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The possibility for a new or different type of accident from any
accident previously evaluated is not created as a result of this
amendment. The changes described in the amendment are supported by
the analyses and evaluations described in Attachment 2 (safety
evaluation). The evaluation of the effects of the proposed changes
indicate that all design standards and applicable safety criteria
limits are met. These changes therefore do not cause the initiation
of any new or different accident nor create any new failure
mechanisms.
All equipment important to safety will continue to operate as
designed. Component integrity is not challenged. The changes do not
result in any event previously deemed incredible being made
credible. The changes do not result in more adverse conditions or
result in any increase in the challenges to safety systems.
Therefore, operation of the Point Beach Nuclear Plant in accordance
with the proposed amendments will not create the possibility of a
new or different type of accident from any accident previously
evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not involve a significant reduction in
a margin of safety.
The proposed changes do not involve a significant reduction in
the margin of safety. Existing component redundancy is not being
changed by these proposed changes. There are no new or significant
changes to the initial conditions contributing to accident severity
or consequences. The margin of safety is maintained by assuring
compliance with acceptance limits reviewed and approved by the NRC.
Since all of the appropriate acceptance criteria for the various
analyses and evaluations have been met as discussed in Attachment 2
(Safety Evaluation) of this submittal and provided for information
in Attachment 4 (PBNP FSAR Chapter 14 ``Safety Analysis'' changes
required as a result of the analyses performed for the upgraded
fuel) of this submittal, by definition there has not been a
significant reduction of any 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.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: July 1, 1999 (TSCR 214).
Description of amendment request: The proposed amendments reflect a
change to Point Beach Nuclear Plant (PBNP) Units 1 and 2 Technical
Specification (TS) Section 15.5.4. The amendment request proposes to
remove one of the two separate methods for verifying the acceptability
of reactor fuel for placement and storage in the spent fuel pool.
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. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes are administrative only in that they remove
the ability to use the reference Koo method for
determining the acceptability of fuel for placement and storage in
the spent fuel pool and new fuel storage vault at the Point Beach
Nuclear Plant. Use of the remaining approved method and requirements
ensure that fuel placed or stored in the spent fuel pool and new
fuel storage vault continues to be in accordance with their
respective design and licensing basis. That is, fuel in the storage
array will continue to meet the design basis requirement that
Keff remain less than 0.95. No modifications are being
made to the spent fuel pool and its cooling system or to the new or
spent fuel storage racks. Since the design basis of the fuel and
storage racks continue to be met, operation in accordance with the
proposed amendments cannot create a significant increase in the
probability or consequences of an accident previously evaluated.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
No physical modifications are being made to the spent fuel pool
and cooling system or to the new or spent fuel storage racks. All
design basis requirements for ensuring the safe storage of fuel in
the spent fuel pool continue to be met. Therefore, operation in
accordance with the proposed amendments cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not create a significant reduction in a
margin of safety.
Technical Specification requirements for placing and storing
fuel in the spent fuel pool continue to ensure that the design basis
requirement, Keff for the fuel array in the spent fuel
pool and new fuel storage remains less than 0.95, is maintained. The
existing margin of safety established by this design requirement is
maintained. Therefore, operation in accordance with the proposed
amendments cannot create a 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.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for license: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Previously Published Notice of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
Texas Utilities Electric Company, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell
County, Texas
Date of amendment request: May 27, 1999, as supplemented by letter
dated May 28, 1999
[[Page 40912]]
Description of amendment request: The proposed amendments would add
a footnote to Technical Specification (TS) 4.8.2.1e, ``D.C. Sources-
Operating,'' which would, on a one-time basis for Unit 1 Battery
BT1ED2, allow the licensee to substitute a performance discharge test
``* * * in lieu of the battery service test required by Specification
4.8.2.1d, twice within a 60 month interval.''
Date of publication of individual notice in Federal Register: June
14, 1999. (64 FR 31881).
Expiration date of individual notice: July 14, 1999.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter 1, 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, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: December 21, 1998, as
supplemented on January 28, February 18, April 2, April 15, and April
16, 1999.
Brief description of amendment: This amendment makes changes to
Facility Operating License No. DPR-35, the Technical Specifications,
and Materials License No. 20-07626-04 to reflect the transfer of the
licenses from Boston Edison Company to Entergy Nuclear Generation
Company.
Date of issuance: July 13, 1999.
Effective date; As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 181.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: January 26, 1999 (64 FR
3984). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 29, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: March 3, 1999.
Brief description of amendment: The amendment modified Technical
Specification Table 4.6-3, ``Reactor Vessel Material Surveillance
Program Withdrawal Schedule.'' The amendment changed the withdrawal
schedule for the upcoming reactor vessel surveillance capsule pull from
approximately 15 effective full power years to approximately 18
effective full power years.
Date of issuance: July 15, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 182.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27316). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 15, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
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: December 1, 1997, supplemented
August 26, 1998.
Brief description of amendments: Revised the Technical
Specifications (TS), Appendix B, Environmental Protection Plan (Non-
Radiological), to implement the terms and conditions of the incidental
Take Statement included in the Biological Opinion issued by the
National Marine Fisheries Service, regarding endangered sea turtles.
Date of Issuance: July 2, 1999.
Effective Date: July 2, 1999.
Amendment Nos.: 162 and 103.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the TS.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68305). The supplemental letter dated August 26, 1998, provided
clarifying information that did not change the original no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New Location County, Connecticut
Date of application for amendment: June 5, 1998, as supplemented
January 13, 1999.
Brief description of amendment: The proposed revision to the
Millstone Unit 3 licensing basis would address a recent steam generator
tube rupture (SGTR) analysis that was determined to be an unreviewed
safety question. The SGTR analyses described in the Final Safety
Analysis Report (FSAR) include an offside dose analysis and a margin to
overfill analysis. Both of the analyses have been updated. The offsite
dose analysis was updated to reflect a larger capacity for the steam
generator atmospheric dump valve (ADV) and a decrease in the operator
response time to close the ADV block valve. The
[[Page 40913]]
margin to overfill analysis was updated to reflect a new single
failure.
Date of issuance: July 2, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 172.
Facility Operating License Nos. DPR-49: Amendments authorizes
revisions to the FSAR,
Date of initial notice in Federal Register: July 1, 1998 (63 FR
35992).
The January 13, 1999, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Northeast Nuclear Energy Company, et al., Docket No. 50-335, Millstone
Nuclear Power Station, Unit No. 2, New Location County, Connecticut
Date of application for amendment: March 19, 1999.
Brief description of amendment: The amendment relocated Technical
Specifications Sections 3.3.3.2, ``Instrumentation, Incore Detectors,''
3.3.3.3, ``Instrumentation, Seismic Instrumentation,'' and 3.3.3.4,
``Instrumentation, Meteorological Instrumentation,'' to the Millstone,
Unit No. 2, Technical Requirements Manual. Index page V and TS Bases
have been revised to reflect the above relocations.
Dated of issuance: July 13, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 237.
Facility Operating License Nos. DPR-65: Amendment Revised the
Technical Specifications.
Date of initial notice in Federal Register: April 21, 1998 (64 FR
19560).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric
Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendment: November 26, 1999, which was
superseded by letter dated June 1, 1998, as supplemented by letters
dated October 30, 1998, March 29, 1999, April 20, 1999, and May 28,
1999.
Brief description of amendment: These amendment would replace the
current ultimate heat sink average water temperature limit for all
combination of plant operations.
Dated of issuance: July 6, 1999.
Effective date: Both units, effective as of date of issuance and
shall be implemented within 30 days.
Amendment Nos.: 182 and 156.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 20, 1998 (63 FR
27764). The October 30, 1998, March 29, 1999, April 20, 1999, and May
28, 1999, letters provided clarifying information that did not change
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 6, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilker-Barre, PA 18701.
PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric
Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendment: November 23, 1999.
Brief description of amendments: The amendments modified the
Susquehanna Steam Electric Station, Units 1 and 2, Technical
Specifications limiting condition for operation, 3.8.3, and
surveillance requirements, 3.8.3.1, to increase the minimum fuel oil
storage tank volume ranges.
Dated of issuance: July 7, 1999.
Effective date: Units 1 and 2, as of date of issuance and shall be
implemented within 30 days.
Amendment Nos.: 183 and 157.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4160).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 7, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference, Department, 71 South Franklin Street, Wikes-Barre, PA 18701.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: October 17, 1997, as
supplemented March 2 and November 28, 1998.
Brief description of amendments: These amendments authorize changes
to the updated Final Safety Analysis Report (FSAR) to permit
installation of digital radiation monitors for both the containment
purge isolation and the control room isolation signals.
Date of issuance: July 12, 1999.
Effective date: July 12, 1999; implementation shall include
submission by the licensee of the revised description authorized by
these amendments with the next update of the FSAR in accordance with 10
CFR 50.71(e).
Amendment Nos.: Unit 2-154; Unit 3-145.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the FSAR.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4324).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of application for amendments: January 12,1999, as
supplemented by letters dated May 11, and June 30, 1999.
Brief description of amendments: The amendment revised Technical
Specification 3/4.7.5, Ultimate Heat Sink, by adding a new action
statement to be used in the event the plant inlet water temperature
exceeds 90 deg. F. The amendment is effective only through September
30, 1999, and is only for the current TSs. The amendment is also
limited to a maximum plant inlet water temperature of 94 deg. F. The
proposal to raise this temperature to 95 deg. F will be addressed in a
future letter.
[[Page 40914]]
Date of issuance: July 8, 1999.
Effective date: July 8, 1999, shall be implemented within 30 days
of the date of issuance.
Amendment No.: 125.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 ( 64
FR 9203). The May 11 and June 30, 1999, 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,
except that the licensee proposed a maximum plant inlet water
temperature of 95 deg. F. where the letters of January and May 11,
1999, proposed only 94 deg. F. The amendment is limited to a maximum
temperature of 94 deg. F.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Dated at Rockville, Maryland, this 21st day of July 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management Office of Nuclear
Reactor Regulation.
[FR Doc. 99-19133 Filed 7-27-99; 8:45 am]
BILLING CODE 7590-01-M
