[Federal Register Volume 64, Number 173 (Wednesday, September 8, 1999)]
[Notices]
[Pages 48858-48875]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-23300]
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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 Public Law 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 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 August 14, 1999, through August 27, 1999.
The last biweekly notice was published on August 25, 1999 (64 FR
46424).
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.
[[Page 48859]]
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 October 8, 1999, 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 2.714 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 must 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.
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.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: August 3, 1999.
[[Page 48860]]
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 2.1.B to increase the minimum
critical power ratio for higher cycle exposures for Unit 2. The
proposed amendments would also revise TS 6.9.A.6.b for Units 2 and 3 to
add an NRC-approved topical report to the list of analytical
methodologies that are used to determine 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, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. These changes do
not affect the operability of plant systems, nor do they compromise
any fuel performance limits.
Changing the Minimum Critical Power Ratio (MCPR) Safety Limit
(SL) at Dresden Nuclear Power Station Unit 2 will not increase the
probability or the consequences of an accident previously evaluated.
This change implements the MCPR SL resulting from the Siemens Power
Corporation (SPC) ANFB critical power correlation methodology using
the approved ATRIUM-9B additive constant uncertainty. For each
cycle, specific MCPR SL calculations will be performed, consistent
with SPC's approved methodology, to confirm the appropriateness of
the MCPR SL. Additionally, operational MCPR limits will be applied
that will ensure the MCPR SL is not violated during all modes of
operation and anticipated operational occurrences. The MCPR SL
ensures that less than 0.1% of the rods in the core are expected to
experience boiling transition. Therefore, the probability or
consequences of an accident will not increase.
Adding EMF-85-74, Revision 0, Supplements 1 and 2 (P)(A) to
Section 6 for Dresden Nuclear Power Station Units 2 and 3, does not
increase the probability or consequences of an accident previously
evaluated. The NRC-approved burnup extension for RODEX2A
applications has been demonstrated to meet all applicable design
criteria. Therefore, adding this methodology to Technical
Specification Section 6 does not increase to the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated:
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications to the plant configuration, including changes in
allowable modes of operation. This Technical Specification submittal
does not involve any modifications to the plant configuration or
allowable modes of operation. No new precursors of an accident are
created and no new or different kinds of accidents are created.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Changing the MCPR SL does not create the possibility of a new
accident from any accident previously evaluated. This change does
not alter or add any new equipment or change modes of operation. The
MCPR SL is established to ensure that 99.9% of the rods avoid
boiling transition.
The MCPR SL is changing for Dresden Nuclear Power Station Unit 2
to support Cycle 17 operation. This change does not introduce any
physical changes to the plant, the processes used to operate the
plant, or allowable modes of operation. Therefore, no new accidents
are created that are different from any accident previously
evaluated.
The addition of RODEX2A (EMF-85-74, Revision 0, Supplements 1
and 2 (P)(A)) to Section 6 does not create the possibility of a new
accident from an accident previously evaluated. This change does not
alter or add any new equipment or change modes of operation. This
change does not introduce any physical changes to the plant, the
processes used to operate the plant, or allowable modes of
operation. Therefore, no new accidents are created that are
different from any accident previously evaluated.
3. Involve a significant reduction in the margin of safety for
the following reasons:
Changing the MCPR SL for Dresden Nuclear Power Station Unit 2
will not involve any reduction in margin of safety. The MCPR SL
provides a margin of safety by ensuring that less than 0.1% of the
rods are calculated to be in boiling transition. The proposed
Technical Specification amendment request reflects the MCPR SL
results from evaluations by SPC using NRC-approved methodology.
Because the methodology used to determine the MCPR SL is
conservative and has received NRC approval, a decrease in the margin
to safety will not occur due to changing the MCPR SL. The revised
MCPR SL will ensure the appropriate level of fuel protection.
Additionally, operational limits will be established based on the
proposed MCPR SL to ensure that the MCPR SL is not violated during
all applicable modes of operation including anticipated operation
occurrences. This will ensure that the fuel design safety criterion
of more than 99.9% of the fuel rods avoiding transition boiling
during normal operation as well as during an anticipated operational
occurrence is met.
The addition of EMF-85-74, Revision 0, Supplements 1 and 2
(P)(A) to Section 6 does not decrease the margin of safety. The
burnup limit extension for RODEX2A applications has been reviewed
and approved by the NRC. The data supporting the burnup extension
demonstrates that all applicable design criteria are met. Therefore,
since the burnup extension is acceptable and within the design
criteria, using the approved burnup extension will not affect 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
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.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: August 13, 1999, as supplemented on
August 27, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification Section 1.0, ``Definitions,'' Item 1.7,
``Core Alteration,'' to specify that movement of instrumentation and
control rod movements are not considered core alterations if there are
no fuel assemblies in the associated cell. The licensee also proposed
corresponding changes to TS Sections 3/4.1, 3/4.3, and 3/4.9 to reflect
the change in definition.
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 incorporate a definition contained in
NUREG-1433, Revision 1, ``Standard Technical Specifications, General
Electric Plants, BWR/4.'' There are no modifications to plant
equipment or systems and there is no direct effect on plant
operation. The proposed changes do not affect any accident
initiators or precursors and do not change or alter the design
assumptions for systems or components used to mitigate the
consequences of an accident. The proposed changes do not affect the
design or operation of any system, structure, or component in the
plant. The proposed changes do not impact
[[Page 48861]]
the requirements for refueling evolutions associated with shutdown
margin, core monitoring, and reactor protection system operability.
There are no changes to parameters governing plant operation, and no
new or different types of equipment will be installed. These changes
do not impact any accident previously evaluated in the Updated Final
Safety Analysis Report (UFSAR). Therefore, no increases in the
probability of an accident or consequences will result due to this
change.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes do not affect the design or operation of
any plant system, structure, or component. There are no changes to
parameters governing plant operation, and no new or different type
of equipment will be installed. There is no change in any method by
which a safety related system performs its function. No new
equipment is being introduced, and installed equipment is not being
operated in a new or different manner. There are no setpoints
affected by this proposed action. This proposed action will not
alter the manner in which equipment operation is initiated, nor will
the function demands on credited equipment be changed. As such, no
new failure modes are being introduced. There are no changes to
assumptions in accident analysis. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed changes are consistent with NUREG-1433, Revision 1,
``Standard Technical Specifications, General Electric Plants, BWR/
4.'' The proposed changes do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. The initial conditions and
methodologies used in the accident analyses remain unchanged.
Therefore, accident analyses results are not impacted. There are no
resulting effects on plant safety parameters or setpoints. The
proposal does not involve a significant relaxation of the criteria
used to establish safety limits, a significant relaxation of the
bases for the limiting safety system settings, or a significant
relaxation of the bases for the limiting conditions for operations.
Therefore, these proposed changes do not cause a reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
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.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: May 5, 1999.
Description of amendment request: The proposed amendment would
permit a one-time extension of the allowed outage time (AOT) for the
reactor protection and engineered safety feature actuation
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The reactor protection and engineered safety features functions
are not initiators of any design basis accident or event and
therefore do not increase the probability of any accident previously
evaluated. The proposed changes to the AOTs, bypass times, and
allowing on-line testing and maintenance have an insignificant
impact on plant safety based on the calculated CDF [core damage
frequency] increase being less than LOE-06. Therefore, the proposed
changes do not result in a significant increase in the consequences
of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not result in a change in the manner in
which the RPS [reactor protection system] and ESFAS [engineered
safety features actuation system] provide plant protection. No
change is being made which alters the functioning of the RPS and
ESFAS. Rather, the likelihood or probability of the RPS or ESF
functioning properly is affected as described above. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident nor involve a reduction in the margin of safety as
defined in the Safety Analysis Report.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not alter the manner in which safety
limits, limiting safety system setpoints or limiting conditions for
operations are determined. The impact of increased AOTs, testing
times, and allowing on-line testing and maintenance are expected to
result in an overall improvement in safety because:
The longer AOTs for the master relays, logic cabinets, and
analog channels will promote improved maintenance practices that
will provide improved component performance, improved availability
of the protection system, and a reduced number of spurious reactor
trips and spurious actuation of safety equipment.
The longer AOTs and bypass times for the analog channels will
provide additional time before being required to place the channel
in trip. With the channel in trip, the logic required to cause a
reactor trip or a safety system actuation is reduced to 1 of 2 (for
2 of 3 logic) and to 1 of 3 (for 2 of 4 logic). With the reduced
logic requirement, the potential for a spurious actuation is
increased. Leaving the channel in the bypass state for additional
time does reduce the availability of signals to initiate component
actuation for event mitigation when required, but as shown in this
analysis, the impact on plant safety is small due to the
availability of other signals or operator action to trip the reactor
or cause component actuation.
The longer allowed outage times will provide plant operators
additional flexibility in operating the plant. There will be
additional time available before an action needs to be taken to shut
down the plant or place a channel in the tripped state. This
additional flexibility will facilitate prioritizing component
repairs.
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.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610. Biweekly Notice
Coordinator Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving
Place, New York, New York 10003.
NRC Section Chief: S. Singh Bajwa.
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: August 4, 1999.
Description of amendment request: The amendments would revise the
joint Technical Specifications as follows:
(1) A current action in Section 3.2.2 requires that when one
Nuclear Service Water System (NSWS) suction transfer low pit level
channel is inoperable, the channel be placed in its trip position. The
licensee proposed an additional alternative such that the NSWS suction
can simply be aligned from Lake Wylie to the Standby Nuclear Service
Water Pond (SNSWP). Suction from Lake Wylie is the normal
configuration, while suction from the SNSWP is the safety
configuration. This proposed alternative
[[Page 48862]]
action provides operational flexibility; there is no associated design
change to the units.
(2) The licensee proposed to delete from Table 3.3.2-1,
``Engineered Safety Feature Actuation System Instrumentation,'' the
entry regarding Auxiliary Feedwater Loss of Offsite Power (Function 6d)
on the basis that a comparable and adequate requirement will exist in
Section 3.3.5. To such end, a new Surveillance Requirement (SR) 3.3.5.3
will be added, incorporating the Function 6d requirement from Table
3.3.2-1. These proposed changes remove inconsistencies that currently
exist in the Technical Specifications for Function 6d. There is no
associated design change to the units.
(3) In the process of converting the Technical Specification to the
improved format (Amendment Nos.173 and 165), errors were inadvertently
introduced regarding the conditions under which the Reactor Coolant
System Subcooling Margin Monitor must be operable. The licensee
proposed to correct these errors by revising the entry regarding the
Subcooling Margin Monitor in Table 3.3.3-1, ``Post Accident Monitoring
Instrumentation''. There is no associated design change to the units.
(4) Section 3.4.17 is concerned with reactor coolant system loops
test exceptions. Currently Surveillance Requirement 3.4.17.2
incorrectly specifies that a COT [channel operational test] be
performed ``for each power range neutron flux-flow and intermediate
range neutron flux channel and P-7 [Low Power Reactor Trips Block
Function]''. The licensee proposed to correct this statement by
deleting ``P-7'' and adding ``P-10 [Power Range Neutron Flux] and P-13
[Turbine Impulse Pressure]''. This correction does not involve any
design change to the units.
(5) The licensee proposed to delete from Section 5.3.1 the specific
qualification requirements for Reactor Operators (ROs) and Senior
Reactor Operators (SROs). Such requirements are specified by 10 CFR
50.55, ``Operators'' Licenses'', and the licensee is required to follow
this regulation. There will be no change in the qualification of ROs
and SROs, and no design change to the units.
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
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Approval of this amendment will have no effect
on accident probabilities or consequences. For proposed changes #1-
4, the systems and equipment referenced in the revised TS are not
accident initiating systems; therefore, there will be no impact on
any accident probabilities by the approval of this amendment. The
design of the systems is not being modified by these proposed
changes. Therefore, there will be no impact on any accident
consequences. For proposed change #5, the change is purely
administrative; it will therefore have no effect on any accident
probabilities or consequences.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant which will introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators; neither does it adversely
impact any accident mitigating systems.
Third Standard
Implementation of this amendment would 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 design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed amendment. The systems and equipment referenced in the
revised TS for proposed changes #1-4 are already capable of
performing as designed. No safety margins will be impacted. Since
proposed change #5 is purely administrative, it will have no effect
on any safety margins.
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: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: July 26, 1999.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 3/4.3.2.1, ``Safety
Features Actuation System Instrumentation,'' to remove the ``Trip
Setpoint'' values and revise the ``Allowable Values'' entries for
Sequence Logic Channels a, ``Essential Bus Feeder Breaker Trip (90%),''
and b, ``Diesel Generator Start, Load Shed on Essential Bus (59%).''
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:
The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the
proposed changes and determined that a significant hazards
consideration does not exist because operation of the Davis-Besse
Nuclear Power Station, Unit No. 1, in accordance with these changes
would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the proposed changes do not
change any accident initiator, initiating condition, or assumption.
The proposed changes would revise Technical Specification (TS)
Table 3.3-4, Safety Features Actuation System Instrumentation Trip
Setpoints, to remove the'Trip Setpoint'' values for Functional Unit
Sequence Logic Channel ``a'', ``Essential Bus Feeder Breaker Trip
(90%)'', and Functional Unit Sequence Logic Channel ``b'', ``Diesel
Generator Start, Load Shed on Essential Bus (59%)'', and also modify
the ``Allowable Values'' entry for Functional Unit Sequence Logic
Channel ``a'', consistent with updated calculations and current
setpoint methodology. The proposed changes would also clarify an
inconsistency between Table 3.3-4 and Table 4.3-2, Safety Features
Actuation System Instrumentation Surveillance Requirements. The
proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1
and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.
The accident previously evaluated in Section 15.2.9, ``Loss of
All AC Power to the Station Auxiliaries (Station Blackout),'' of the
DBNPS Updated Safety Analysis Report (USAR) is not affected by the
proposed changes because its bounding conditions are not affected.
The existing TS action statements will continue to maintain the USAR
requirement to start and load one Emergency Diesel Generator (EDG)
to meet minimum ESF requirements, should all AC power be lost.
Furthermore, the proposed changes are based on the existing
performance characteristics of plant equipment; therefore, the
proposed changes
[[Page 48863]]
will not involve a significant change to the plant design or
operation.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate assumptions used in evaluating the radiological
consequences of an accident, do not alter the source term or
containment isolation, and do not provide a new radiation release
path or alter radiological consequences.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not introduce a new or different accident initiator or
introduce a new or different equipment failure mode or mechanism.
3. Not involve a significant reduction in a margin of safety
because the proposed changes do not significantly reduce the ability
of the plant to respond to a loss of AC power to the essential 4160
Volt buses in a timely manner. The revised Allowable Value for the
Sequence Logic Channel ``Essential Bus Feeder Breaker Trip (90%)''
takes into account the need not only to be able to actuate
Engineered Safety Features equipment coincident with a degraded grid
condition, but to provide voltage at the required value to properly
operate the equipment.
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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: July 27, 1999.
Description of amendment request: The proposed amendment would
remove Technical Specification (TS) Section 6.4, ``Training,'' relocate
TS Sections 6.5.2.8, ``Audits,'' and 6.10 ``Record Retention,'' to the
Updated Safety Analysis Report, and make related changes to TS Sections
6.14, ``Process Control Program,'' and 6.15, ``Offsite Dose Calculation
Manual.'' In addition, an editorial correction is proposed to TS 6.8,
``Procedures and Programs.''
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:
The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power
Station, Unit Number 1, in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions or assumptions are affected by the proposed changes to
Section 6.0, Administrative Controls, of the Technical
Specifications (TS).
The proposed changes to remove Section 6.4, Training, from the
TS and relocate the detailed listings of TS Section 6.5.2.8, Audits,
and TS Section 6.10, Record Retention, to the DBNPS [Davis-Besse
Nuclear Power Station] Quality Assurance Program in Chapter 17 of
the Updated Safety Analysis Report are consistent with NUREG-1430,
``Standard Technical Specifications--Babcock and Wilcox Plants,''
Revision 1 or NRC Administrative Letter 95-06 ``Relocation of
Technical Specification Administrative Controls Related to Quality
Assurance,'' dated December 12, 1995. The proposed changes to TS
Section 6.14, Process Control Program (PCP); TS Section 6.15,
Offsite Dose Calculation Manual (ODCM); and TS Section 6.8,
Procedures and Programs, are either associated administratively with
the above proposed changes or are editorial corrections. These TS
being removed or relocated will remain subject to the controls of
regulations (e.g., 10 CFR 50.59, 10 CFR 55.59, or 10 CFR 50.54(a)).
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident conditions or
assumptions are affected by the proposed changes. As described
above, these changes are consistent with the improved ``Standard
Technical Specifications--Babcock and Wilcox Plants'' (NUREG-1430)
or Administrative Letter 95-06 and are administrative changes. The
proposed changes do not alter the source term, containment
isolation, or allowable releases. The proposed changes, therefore,
will not increase the radiological consequences of a previously
evaluated accident.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes, which involve only administrative controls. The proposed
changes do not alter any accident scenarios.
3. Not involve a significant reduction in a margin of safety
because the proposed changes are administrative and do not reduce or
adversely affect the capabilities of any plant structures, systems
or components to perform their nuclear safety function.
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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: August 5, 1999.
Description of amendment request: The requested changes correct
editorial errors in Technical Specification (TS) Sections 3.8.3.2,
4.6.2.1, 4.6.2.2, 4.8.1.1, and 4.9.12. Also, the requested changes
correct minor editorial and reference errors in Technical Specification
Bases Sections B 3/4.3.2, B 3/4.4.11, B 3/4.6.1.2, and B 3/4.8.4.
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:
NNECO [Northeast Nuclear Energy Company] has reviewed the
proposed revision in accordance with 10CFR50.92 and has concluded
that the revision does not involve any Significant Hazards
Considerations (SHC). The basis for this conclusion is that the
three criteria of 10CFR50.92(c) are not satisfied. The proposed
Technical Specification revision does not involve an SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed TS changes are editorial in nature and do not alter
or effect the design, operation, maintenance[,] or surveillance
associated with MP-3 [Millstone Nuclear Power Station, Unit No. 3]
[s]tructures, [s]ystems, and [c]omponents (SSC) during normal or
accident operations. Since the SS[Cs] are not altered[,] the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed TS changes are editorial in nature and do not alter
or effect the design,
[[Page 48864]]
operation, maintenance[,] or surveillance associated with MP-3
[s]tructures, [s]ystems, and [c]omponents (SSC) during normal or
accident operations. Since the Units SS[Cs] have not been modified
physically, or operationally[,] due to procedure changes prompted by
this TSCR [Technical Specification Change Request], the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Involve a significant reduction in the margin of safety.
These proposed TS changes are editorial and do not impact any
MP-3 design or operational requirements. MP-3 system performance and
operating limits are not affected; therefore[,] the proposed change
does not involve a significant reduction in the margin of safety.
In conclusion, based on the information provided, it is
determined [by NNECO] that the proposed revision does not involve
a[n] SHC.
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: 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 22, 1999.
Description of amendment request: The Limerick Generating Station
(LGS), Units 1 and 2, Technical Specifications (TS) contained in
Appendix A to the Operating Licenses would be amended to eliminate a
surveillance requirement for the Reactor Recirculation System. This
proposed TS change request involves revising the TS to delete
Surveillance Requirement 4.4.1.1.2, and associated TS Administrative
Controls Section 6.9.1.9.h, which requires that each Reactor
Recirculation System pump motor generator (MG) set scoop tube
mechanical and electrical stop be demonstrated OPERABLE with the
overspeed setpoints less than or equal to the setpoints as noted in the
Core Operating Limits Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of an
accident previously evaluated. The proposed TS changes do not make any
physical changes to the fuel, or the way the fuel responds to a
transient or accident. The radiological barriers are not compromised.
The fuel will continue to be operated to analyzed operating limits. No
new failure mode is introduced.
Prior to the removal of the Recirculation System Master Flow
Controller at LGS, the bounding postulated event involving an increase
in reactor coolant system flow rate was the dual pump slow flow runout
event not terminated by SCRAM. The requirements surrounding the MG set
stops were established to mitigate consequences during a dual pump slow
flow runout by providing a limit on the maximum core flow. The MG set
stop requirements were not established to prevent an accident. The
potential common mode failure required for a dual pump slow flow runout
event was eliminated with the removal of the Master Flow Controller.
The elimination of the Master Flow Controller does not increase the
probability of other core flow increase events, or of any other events
previously analyzed.
Revised generic flow biased ARTS [APRM (average power range
monitor)/RBM (rod block monitor) Technical Specifications Improvement]
thermal limits that do not take credit for MG set stops have been
developed for LGS, Units 1 and 2. Adherence to approved flow biased
ARTS thermal limits identified in the LGS, Units 1 and 2, Core
Operating Limits Reports (COLRs) ensure that fuel design limits are not
exceeded. Maintaining fuel design limits results in no change in the
consequences of accidents previously evaluated.
The single pump slow flow runout does not terminate by Main Steam
Isolation Valve (MSIV) closure or generator load reject. As a result,
the single pump runout event does not result in any significant
pressurization and does not represent a challenge to the reactor
coolant pressure boundary. MSIV closure with associated SCRAM on high
neutron flux, as confirmed in the cycle specific Supplemental Reload
Licensing Report (SRLR), remains the bounding reactor pressure vessel
overpressurization event for LGS, Units 1 and 2. In addition, there are
no other associated impacts to the plant resulting from a single pump
runout. Therefore, the integrity of radiological barriers will not be
compromised.
Although there is no longer a safety need to demonstrate
operability of the MG set stops, there still is an operational need to
have the MG set stops for the Reactor Recirculation System (RS). Damage
to the jet pump sensing lines could occur if the resonance frequency of
the sensing lines is reached. Jet pump sensing line tests established a
conservative pump speed limit (1650 rpm for Unit 1, no limit for Unit
2) to preclude sensing line resonance. The MG set stop setpoint bounded
the operationally required setpoint. The operationally required MG set
stop setpoint to preclude jet pump sensing line resonance will continue
to be controlled administratively via approved plant procedures. The
proposed TS changes do not adversely impact the RS, or introduce new or
unanalyzed operating conditions for the RS. The MG sets will not exceed
their previously analyzed maximum 57.5 Hz with the stops removed.
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed TS changes do not make any physical changes to the fuel,
or the way the fuel responds to a transient or accident. The
radiological barriers are not compromised. The fuel will continue to be
operated to analyzed operating limits. No new failure mode is
introduced.
The proposed TS changes do not create new operating conditions that
have not been evaluated. Removal of the Recirculation Master Flow
Controller eliminates the possibility of a single failure initiated
common mode event. Since the possibility of a common failure has been
eliminated, the most limiting recirculation runout event is a one pump
slow flow runout. This is the same kind of postulated accident as that
previously evaluated, only it involves one pump instead of both pumps.
Therefore, the proposed TS changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
[[Page 48865]]
3. The proposed TS changes do not involve a significant reduction
in a margin of safety.
The proposed TS changes do not make any physical changes to the
fuel, or the way the fuel responds to a transient or accident. The
radiological barriers are not compromised. The fuel will continue to be
operated to analyzed operating limits. No new failure mode is
introduced.
Single pump runout based, generic flow biased ARTS thermal limits
that do not take credit for MG set stops have been developed for LGS,
Units 1 and 2. Adherence to approved ARTS-based flow biased thermal
limits identified in the LGS, Units 1 and 2, COLRs and implemented in
the plant process computer are sufficient to maintain the margin of
safety as delineated in TS Sections 3/4.2.1, 3/4.2.3, and 3/4.2.4.
Therefore, these proposed TS changes do not involve a significant
reduction in a margin of safety.
Based on the above review, the NRC staff concludes that 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of amendment request: January 29, 1998.
Description of amendment request: The amendment would delete the
requirements for a security plan from the 10 CFR Part 50 license and
technical specifications after the spent nuclear fuel is transferred to
a Part 72 licensed independent spent fuel storage installation (ISFSI).
Security requirements for the ISFSI would be in accordance with 10 CFR
Part 72, Subpart H.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The physical structures, systems and components of the Trojan
Nuclear Plant and the operating procedures for their use are
unaffected by the proposed change. The proposed elimination of the
security requirements for the 10 CFR Part 50 license, is predicated
on approval of the Trojan ISFSI Security Plan (PGE 1073) which will
be coincident with issuance of a 10 CFR Part 72 license and upon
completion of the transfer of all nuclear fuel from the spent fuel
pool to the ISFSI. The planned 10 CFR 72 licensing controls for the
ISFSI will provide adequate confidence that personnel and equipment
can perform satisfactorily for normal operations of the ISFSI and
respond adequately to abnormal events/accidents. The proposed Trojan
ISFSI Security Plan (PGE 1073) will also provide confidence that
security personnel and safeguards systems will perform
satisfactorily to ensure adequate protection for the storage of
spent nuclear fuel. Therefore, the proposed 10 CFR Part 50 amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change is security related, and as such, has no
direct impact on plant equipment or the procedures for operating
plant equipment and, therefore, does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. Because the proposed ISFSI area will be segregated from
the 10 CFR Part 50 licensed area, licensed security activities under
the 10 CFR Part 50 license will no longer be necessary after all the
nuclear fuel has been moved. The planned 10 CFR 72 licensing
controls for the ISFSI area will provide adequate confidence that
personnel and equipment can perform satisfactorily for normal
operations of the ISFSI and respond adequately to normal events/
accidents. Moreover, the ISFSI will be physically separate from the
Trojan Nuclear Plant structures and equipment. Therefore, the
proposed 10 CFR Part 50 license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The assumptions for a fuel handling and other accidents are not
affected by the proposed license amendment. Because the proposed
ISFSI area (that will contain the nuclear fuel) will be segregated
from the 10 CFR Part 50 licensed area, licensed security activities
under the 10 CFR Part 50 license will no longer be necessary. The
planned 10 CFR 72 licensing controls for the ISFSI area will provide
adequate confidence that personnel and equipment can perform
satisfactorily for normal operations of the ISFSI and respond
adequately to abnormal events/accidents. Also, the ISFSI will be
physically separate from the Trojan Nuclear Plant structures and
equipment. Therefore, the proposed 10 CFR Part 50 license amendment
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.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensee: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRC Section Chief: Michael T. Masnik.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: August 19, 1999. The August 19, 1999,
submittal supersedes the February 18, 1999, submittal in its entirety
(64 FR 14284).
Description of amendment request: The proposed amendment would
revise the Virgil C. Summer Nuclear Station (VCSNS) Technical
Specifications (TS) to incorporate the new Pressure/Temperature (P-T)
Limits Curves consistent with the analysis results of reactor vessel
specimen W. These figures are contained in Section 3/4.4.9 and are
presented as Figures 3.4-2 and 3.4-3. These figures were developed
using the methodology included in WCAP 14040-NP-A, ``Methodology Used
to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup
and Cooldown Limit Curves,'' as well as Code Case N-640, ``Alternative
Reference Fracture Toughness for Development of P-T Limit Curves for
Section XI, Division I.'' A reduced flange temperature requirement was
included in the development of the curves, with justification provided
in WCAP 15102, Revision 1, ``V. C. Summer Unit I Heatup and Cooldown
Limit Curves for Normal Operation.'' Additionally, the Bases section
for the Pressure/Temperature Limits would be revised to accurately
reflect current industry standards and regulations. A significant
portion of this Bases section would be deleted due to the information
also being located in WCAP 15102, Revision 1.
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
[[Page 48866]]
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes revise the Pressure/Temperature Limits
Curves to provide curves that reflect the results of the analysis
performed on reactor vessel surveillance specimen W. This analysis
was performed using NRC approved methodology as documented in WCAP
14040-NP-A, utilizing the 1996 ASME Boiler and Pressure Vessel Code,
Section XI, Appendix G requirements, along with ASME Code Case N-
640. These curves provide the limits for operation of the Reactor
Coolant System during heat up, cool down, criticality, and
hydrotesting. These curves are provided without instrument
uncertainties included, however, the uncertainties are included in
the curves provided in the operating procedures. The limits protect
the reactor vessel from brittle fracture by separating the region of
acceptable operation from the region where brittle fracture is
postulated to occur. Failure of the reactor vessel is not a VCSNS
design basis accident, and, in general, reactor vessel failure has a
low probability of occurrence and is not considered in the safety
analysis. Therefore, the change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes revise the Pressure/Temperature Limits
Curves, Section 3/4.4.9, to incorporate the results of the analysis
performed on reactor vessel specimen W. There are no plant design
changes or significant changes in any operating procedures. This
change adjusts the heatup and cooldown curves to reflect the shift
in nil-ductility reference temperature of the reactor vessel as a
result of neutron embrittlement, and alternate methodology utilized
to generate the curves. Therefore, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does this change involve a significant reduction in margin of
safety?
The proposed changes revise the Pressure/Temperature Limits
Curves, Section 3/4.4.9, to incorporate the results of the analysis
performed on reactor vessel specimen W. The new PT curves ensure
that the 10 CFR 50 Appendix G, requirements are not exceeded during
normal operation including Reactor Coolant System transients during
heat up, cool down, criticality, and hydrotesting. The new PT curves
were prepared, using accepted industry methodology, for a projected
reactor vessel neutron exposure of 32 EFPY [Effective Full Power
Years].
The new curves will serve as the basis for operating
limitations, to provide margin against non-ductile fractures. The
uncertainties introduced by instrumentation, forced flow and
elevation differences are not reflected in the TS curves. These
uncertainties will be factored into the curves presented in the
operating procedures. Since administrative limits remain in place to
ensure that 10 CFR 50 Appendix G limits are not challenged, the
margin of safety described in the TS Bases is not reduced by the
proposed change. Therefore, the 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.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: Richard L. Emch, Jr.
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 amendment requests: August 11, 1999 (PCN-488).
Description of amendment requests: The proposed amendments would
modify the Technical Specifications for the San Onofre Nuclear
Generating Station (SONGS) Units 2 and 3 to revise Surveillance
Requirement (SR) 3.3.7.3 by providing allowable values in place of
analytical limits for certain degraded voltage parameters, and by
deleting unnecessary parameter limits in cases where plant safety is
not affected. The proposed change would also delete redundant SR
3.3.7.4.
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.
Proposed Change Number (PCN)-488 revises the Technical
Specification (TS) Surveillance Requirement (SR) acceptance criteria
of the Loss of Voltage Signal (LOVS), Degraded Grid Voltage with
Safety Injection Actuation Signal (DGVSS), and Sustained Degraded
Voltage Signal (SDVS) relay circuits. These circuits are not
accident initiators.
PCN-488 revises the TS SR acceptance requirements to make them
more limiting than the present requirements. Because the revised
acceptance criteria are more limiting than the present requirements,
the consequences of accidents analyzed in the Updated Final Safety
Analysis Report (UFSAR) are not increased. PCN-488 also revises the
TS SR acceptance requirements to delete upper and lower bounds in
cases where the deleted bound provides no safety benefit. Deleting
bounds having no safety significance does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
PCN-488 deletes redundant SR 3.3.7.4, which is not in NUREG-
1432, Standard Technical Specifications, Combustion Engineering
Plants. Deleting a redundant requirement does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Consequently, the proposed amendment does not result in an
increase in the probability of accidents evaluated in the UFSAR.
2. Does this amendment request create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No.
PCN-488 revises the TS SR acceptance criteria of the LOVS,
DGVSS, and SDVS relay circuits, which are not accident initiators,
and deletes a redundant SR. PCN-488 does not introduce any revision
in the hardware configuration of the protective circuitry for LOVS,
DGVSS or SDVS. The measurement required by the deleted, redundant
surveillance is required elsewhere in the TS. For these reasons,
PCN-488 does not create the possibility of any new or different kind
of accident from any previously evaluated. '
3. Does this amendment request involve a significant reduction
in a margin of safety?
No.
PCN-488 provides allowable values for the acceptance criteria
for the TS SR for LOVS, DGVSS and SDVS. As such, the revised values
are more limiting than the current values, which represent design
limits. Therefore, PCN-488 does not involve a significant reduction
in a margin of safety.
PCN-488 also revises the TS SR acceptance requirements to delete
upper and lower bounds in cases where the deleted bound provides no
safety benefit. Deleting bounds having no safety significance does
not involve a significant reduction in a margin of safety.
PCN-488 additionally deletes a redundant SR. Because the deleted
surveillance is required elsewhere in the TS, this action does not
involve a significant reduction in a margin of safety.
For these reasons, PCN-488 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 requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
[[Page 48867]]
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
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 31, 1998, as supplemented by
letters dated April 19 and August 18, 1999. The August 31, 1998,
application was originally noticed in the Federal Register on October
21, 1998 (63 FR 56260).
Description of amendment request: The proposed amendments would
revise Technical Specification 3/4.4.9.3 by revising the cold
overpressure mitigation curve to accommodate the replacement steam
generators and by adding two surveillances (for the centrifugal
charging pumps and the emergency core cooling system accumulators) to
ensure the operability of the cold overpressure mitigation system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Reanalysis of STP [South Texas Project, Units 1 and 2] COMS
[cold overpressure mitigation system] transients to consider design
characteristics of Delta-94 RSGs [replacement steam generators] has
shown that maximum allowable PORV [power-operated relief valve]
setpoints decrease slightly, and continue to provide design basis
low temperature overpressure protection with Delta-94 steam
generators. This change request incorporates the new COMS curves
into Technical Specification 3.4.9.3 (Figure 3.4-4). Maximum
allowable PORV setpoints decrease with Delta-94 steam generators,
and are conservative compared to Model E steam generator curves. Use
of the new curves with either Model E or Delta-94 steam generators
conforms to the STP design basis.
These changes are based on a reanalysis that accounts for Model
Delta-94 design, a decision to make calculation[s] of COMS maximum
allowable PORV setpoint consistent with current industry standards
as represented by WCAP-14040, and addition of two surveillances to
the Technical Specification to ensure operability of COMS. Moving
maximum allowable PORV setpoints in the conservative direction and
adding surveillances to reinforce standard operating practice have
no adverse effect on the probability or consequences of an accident
previously evaluated. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed PORV maximum allowable setpoint changes do not
create any new operating conditions or modes, and the added
surveillances have no effect except to ensure operation of COMS as
designed. The slight change to the maximum allowable PORV setpoint
curves for the Cold Overpressure Mitigation System accommodates
Delta-94 steam generator design characteristics, and COMS continues
to perform in accordance with existing requirements, which are
sufficient to ensure plant safety is preserved.
The proposed change is the result of a reanalysis of a
previously evaluated accident. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change reflects design characteristics of the new
Delta-94 steam generators. The change to the COMS curves is in the
conservative direction and does not affect any design failure point
or system limitation. Therefore, the 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas
77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
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: August 18, 1999.
Description of amendment request: The licensee proposed changing
the Vermont Yankee Nuclear Power Station (VY) Technical Specifications
by revising the reactor core spiral reloading pattern such that it
begins around a source range monitor rather than from the center of the
core. The offloading pattern would be the reverse sequence.
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.
VY has determined that the proposed change to reload the reactor
core in a spiral pattern beginning around a Source Range Monitor
(SRM) does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The design basis
accident associated with refueling is the Refueling Accident; i.e.,
the accidental dropping of a fuel bundle onto the top of the core.
There is no assumption as to the core loading pattern in the
analysis of this accident. The analyzed abnormal operational
transients associated with refueling are: (1) the Control Rod
Removal Error During Refueling, and (2) the Fuel Assembly Insertion
Error During Refueling. There is no assumption as to the core
loading pattern in the analyses of these transients. The Fuel
Assembly Insertion Error During Refueling transient involves
mislocated and rotated fuel assembly loading errors. However, a
change in the approved core loading pattern has no impact on the
probability of mislocating or rotating a bundle while following that
pattern. Furthermore, the proposed change implements a core loading
pattern that provides improved flux monitoring as compared to the
pattern prescribed by the current Technical Specifications. When
loading the core in accordance with the proposed change, the SRM
indication will be indicative of the true flux of the loaded fuel,
as the creation of flux traps (moderator filled cavities surrounded
on all sides by fuel) is precluded.
The SRMs and the core loading pattern are not initiators of any
accident previously evaluated. As such, the subject changes cannot
affect the probability of an accident previously evaluated. The core
loading pattern is not assumed in the mitigation of any accident.
Since the proposed change provides improved flux monitoring by the
SRMs, operators will have more accurate indication and SRM automatic
trip functions will actuate based on a more accurate indication of
flux. As such, any event mitigation function provided by the SRMs is
enhanced by this change. Therefore, the associated changes do not
involve a significant increase in the 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.
VY has determined that the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. VY proposes to change the core reloading and
offloading patterns to start and stop, respectively, at an
[[Page 48868]]
SRM versus the geometric center of the core as prescribed by current
Technical Specifications. This ensures that flux monitoring
instrumentation is always OPERABLE in the fueled region of the
vessel. There is no separation of the monitoring device from the
fuel by cavities of water as is the case with the pattern prescribed
by the current Technical Specifications. As such, flux monitoring is
enhanced during core reloading and offloading. This change is
conservative relative to the current requirements. Therefore, no new
or different kinds of accidents are created.
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.
VY has determined that the proposed change does not involve a
significant reduction in a margin of safety. Loading around the
geometric center of the core as prescribed by the current Technical
Specifications results in cells of moderator separating the fuel
from the instrumentation monitoring its flux. This change requires
the flux monitoring instrumentation to be in the fueled region, and,
in so doing, provides for more accurate monitoring of core flux
during core reloading and offloading. As such, the operators will
have more accurate indication and SRM automatic trip functions will
actuate when the actual flux reaches the trip setpoints. Therefore,
this 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.
Attorney 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: August 18, 1999.
Description of amendment request: The licensee proposed changing
the Vermont Yankee Nuclear Power Station (VY) technical specifications
(TSs) by revising the definition of the ``Surveillance Frequency'' to
incorporate provisions that apply upon the discovery of a missed TS
surveillance. The provisions would allow 24 hours to perform the
surveillance before the applicable limiting condition for operation is
entered.
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.
This change does not result in any physical alteration of plant
systems, structures or components; nor does the change modify the
manner in which plant equipment will be operated or maintained. As a
result, the proposed change does not affect any of the parameters or
conditions that contribute to the initiation or mitigation of any
accidents previously evaluated.
Surveillance frequencies are not assumed in the initiation of
any analyzed event. Thus, conditions assumed in the plant accident
analyses are unchanged. Furthermore, there is no relaxation of
required setpoints or operating parameters.
Therefore, the probability or consequences of an accident
previously evaluated are not significantly increased since the most
likely outcome of performing a surveillance is that it does, in
fact, demonstrate the system or component is operable. VY has,
therefore, determined that the proposed change does not involve 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 change will not modify the
physical plant or the modes of plant operation. The changes do not
involve the addition or modification of equipment nor do they alter
the design or operation of plant systems. These changes to Technical
Specifications do not create any new or different kind of accident
since they do not involve any change to the plant or the manner in
which it is operated.
Therefore, VY has determined that the proposed change does not
create the possibility of a new or different kind of accident from
any accident 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 proposed change does not affect design margins or
assumptions used in accident analyses. The capability of safety
systems to function and limiting safety system settings are
similarly unaffected as a result of this change.
The increased time allowed (up to 24 hours) for the performance
of a surveillance discovered to have not been performed, is
acceptable based on the small probability of an event requiring the
associated component. The requested allowance will provide
sufficient time to perform the missed surveillance in an orderly
manner. Without the 24 hour delay, it is possible that the missed
surveillance would force a plant shutdown; thus, the plant could be
shutting down while the missed surveillance is being performed. As a
result of this delay, the potential for human error will be reduced.
Consequently, there is no significant reduction in a margin of
safety as overall plant safety is enhanced due to the avoidance of
unnecessary plant shutdowns.
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
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia
Date of amendment request: August 4, 1999.
Description of amendment request: The proposed changes to North
Anna Power Station (NAPS) Units 1 and 2 Technical Specification (TS)
4.4.1.6.1 and associated Bases will extend the drained reactor coolant
loop verification time (verified as drained) from two hours to four
hours prior to backfilling when returning the drained loop to service.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated.
Administrative procedures ensure that the initiation of seal
injection in order to establish a partial vacuum in an isolated and
drained loop will not create the potential for an inadvertent and
undetected introduction of under-borated water into an isolated loop
prior to returning the isolated loop to service. Additionally,
extension of the drained loop verification time from two hours to
four hours prior to backfill operations will not significantly
diminish confidence that the isolated and drained loop will, in
fact, be drained at the time the back-fill evolution is initiated.
Therefore, there is no measurable increase in the probability or
consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated.
[[Page 48869]]
There are no modifications to the plant as a result of the
changes. No new accident or event initiators are created by the
initiation of seal injection in order to establish a partial vacuum
in an isolated and drained loop, and by the extension of the drained
loop verification time requirement from two hours to four hours
prior to backfill operations. Therefore, the proposed changes do not
create the possibility of any accident or malfunction of a different
type previously evaluated.
3. Does the change involve a significant reduction in the margin
of safety.
The proposed changes have no effect on the safety analyses
assumptions. Changes acknowledge the establishment of seal injection
for the Reactor Coolant Pump in the isolated and drained loop as a
prerequisite for the vacuum-assisted back-fill technique and extends
the drained-loop verification time from two hours to four hours
prior to backfill operations. The two hour interval was established
to ensure that the drained loop is verified to be drained at a point
in time sufficiently close to the initiation of the back-fill
evolution such that no intervening event could occur that would
render the loop no longer drained. Relaxation of the drained loop
verification time from two hours to four hours will not
significantly diminish confidence that the isolated and drained loop
will be drained at the time the back-fill evolution is initiated.
Therefore, the proposed changes do not result in 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 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 Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Section Chief: Richard L. Emch, Jr.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: April 28, 1999.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) Section 3.4.A.4 and Table 4.1-
2B for Units 1 and 2. The proposed changes would reduce the minimum
volume requirement for the refueling water chemical addition tank (CAT)
to provide additional operating margin, and also correct administrative
format errors in Table 4.1-2B.
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:
Criterion 1--Does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The probability or the consequences of an accident previously
evaluated are not increased. When the revised Safety Analysis Limit
minimum CAT volume of 3800 gallons was implemented, consideration
was given to the effects of the proposed reduced CAT volume on
containment integrity analyses, containment spray and post-LOCA sump
pH analyses, and the post-LOCA recirculation switchover time
interval specified in Emergency Operating Procedures. The change was
determined to be acceptable as accident analyses assumptions would
continue to be met. The proposed TS minimum CAT volume (3930
gallons) includes an allowance for the CAT level Channel Statistical
Allowance (CSA), so that the safety analysis limit CAT volume (3800
gallons) will not be violated when the measured CAT volume (i.e.,
tank level) is at or above the TS minimum CAT volume limit. The
proposed reduction in the TS minimum CAT volume has no bearing on
the probability of occurrence of any accident previously evaluated,
since neither the volume nor the sodium hydroxide inventory of the
CAT have any bearing on postulated accident initiators. Furthermore,
because the affected accident analyses have been evaluated and found
to meet their acceptance criteria with the reduced safety analysis
limit CAT volume, the consequences of an accident previously
evaluated is not increased.
Criterion 2--Does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The possibility of a new or different kind of accident than any
accident previously evaluated is not created. The proposed reduction
in the TS minimum CAT volume does not involve any alterations to the
physical plant that would introduce any new or unique operational
modes or accident precursors. Only the TS minimum CAT volume is
being changed to establish an operationally feasible alarm setpoint
to provide the operators additional flexibility in maintaining the
required CAT volume.
Criterion 3--Does not involve a significant reduction in a
margin of safety.
The margin of safety is not reduced. It was determined that the
affected safety analyses continue to meet their respective
acceptance criteria with the revised minimum CAT volume. By
implementing the proposed change in the TS minimum CAT volume, a CAT
level alarm setpoint may be established which includes a
conservative allowance for level measurement uncertainty such that
neither the proposed TS minimum CAT volume nor the Safety Analysis
Limit CAT volume will be violated at the time a CAT level alarm is
received. Therefore, it is concluded that the proposed change will
not reduce the margin of safety.
This analysis demonstrates that the proposed amendment to the
Surry Units 1 and 2 Technical Specifications does not involve a
significant increase in the probability or consequences of a
previously evaluated accident, does not create the possibility of a
new or different kind of accident and 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.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Section Chief: Richard L. Emch, Jr.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 29, 1998, as supplemented by
letter dated July 29, 1999. The December 29, 1998, amendment
application was previously noticed in the Federal Register on February
24, 1999 (64 FR 9023).
Description of amendment request: The amendment would revise
Section 5.6.6, ``Reactor Coolant System (RCS) Pressure and Temperature
Limits Report (PTLR),'' of the improved Technical Specifications (TSs),
that were issued in Amendment 123 on March 31, 1999. The amendment
would (1) add the phrase ``and Cold Overpressure Mitigation System'' to
the first sentence of item 5.6.6.b that identifies the limits that can
be determined by the licensee in the PTLR, and (2) replace the current
list of documents listed in item 5.6.6.b by the NRC letter that will
approve this amendment and the Westinghouse report, WCAP-14040-NP-A,
``Methodology Used to Develop Cold Overpressure Mitigation System
Setpoints and RCS Heatup and Cooldown Limit Curves,'' dated January
1996. WCAP-14040-NP-A is the NRC-approved topical report that provides
a methodology for developing the cold overpressure mitigation system
(COMS) setpoints and RCS heatup and cooldown limit curves for
Westinghouse plants,
[[Page 48870]]
such as Wolf Creek Generating Station (WCGS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Incorporating the revised heatup and cooldown pressure/
temperature limit curves and the COMS PORV setpoint limit curve into
the WCGS Technical Specifications does not affect the probability or
consequences of an accident previously evaluated.
The revised limit curves are calculated using the most limiting
RTNDT for the reactor vessel components and include a
radiation-induced shift corresponding to the end of the period for
which the curves are generated. The COMS PORV Setpoint Limit Curve
is calculated using the most limiting mass injection transient,
taking into account operation of the NCP [normal charging pump]
during shutdown modes. The changes do not affect the basis,
initiating events, chronology, or availability/operability of safety
related equipment required to mitigate transients and accidents
analyzed for WCGS.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Adopting the revised limit curves redefines the range of
acceptable operation for the Reactor Coolant System. This
redefinition is a result of the analysis of reactor vessel
surveillance specimens removed from the reactor in a continuing
surveillance program which monitors the effects of neutron
irradiation on the WCGS reactor vessel materials under actual
operating conditions. Included in the revised limit curves is
consideration for NCP operation during shutdown modes. Incorporating
these revised curves does not create the possibility of an accident
of a different type from any previously evaluated for WCGS.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The revision of these limit curves continues to maintain the
margin of safety required for prevention of non-ductile failure of
the WCGS reactor vessel during low temperature operation as required
by 10 CFR 50, Appendices G and H. The revised curves primarily
affect RCS operation below 350 deg.F by limiting the available
pressure/temperature window for heatup and cooldown. The revised
limit curves compensate for the in-service radiation induced
embrittlement of the reactor vessel and accounts for the requirement
that the closure flange region temperature must exceed the nil-
ductility temperature by at least 120 deg.F when pressure exceeds
20% of the preservice hydrostatic test pressure.
The revised COMS PORV Setpoint Limit Curve, which includes
consideration of NCP operation during shutdown modes, ensures
overpressure protection of the RCS and reactor vessel.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notice of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notice was previously published as a separate
individual notice. The notice content was the same as above. It was
published as an individual notice either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. It is 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.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of amendment request: August 6, 1999.
Brief description of amendment request: The proposed amendments
would revise the Technical Specifications (TSs) contained in Appendix A
to the Operating Licenses to incorporate a note into the TSs which will
permit a one-time exemption, until September 30, 1999, from the
90 deg.F limit stated in Surveillance Requirement (SR) 3.7.2.2. This SR
currently requires that the average water temperature of the normal
heat sink be less than or equal to 90 deg.F as demonstrated on a 24-
hour frequency. As stated in the proposed TS note, during the time
period between approval and September 30, 1999, the average water
temperature of the normal heat sink will be limited to less than or
equal to 92 deg.F.
Date of publication of individual notice in Federal Register:
August 13, 1999 (64 FR 44243).
Expiration date of individual notice: 14 days for comments, August
27, 1999; 30 days for hearing, September 13, 1999.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
[[Page 48871]]
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: July 30, 1999.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit a 72-
hour delay in the UHS temperature restoration period prior to entering
the plant shutdown required actions. This TS amendment is given as a
temporary amendment change effective until September 30, 1999, after
which the TS will revert back to the original TS provisions.
Date of issuance: August 24, 1999.
Effective date: August 24, 1999.
Amendment No.: 184.
Facility Operating License No. DPR-23: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (64 FR 43406 dated August 10, 1999). The
notice provided an opportunity to submit comments on the Commission's
proposed NSHC determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by September 8,
1999, 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, and final determination of NSHC are contained in
a Safety Evaluation dated August 24, 1999.
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.
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 application for amendments: March 25, 1999.
Brief description of amendments: The amendments revise various
parts of the Technical Specifications (Appendix A of the Catawba
operating licenses) to identify that the Trip Setpoints for the reactor
trip system and engineered safety feature actuation system
instrumentation are in reality Nominal Trip Setpoints.
Date of issuance: August 13, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment Nos.: 179--Unit 1; 171--Unit 2.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24195).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room Location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania
Date of application for amendment: June 18, 1996, as supplemented
December 12, 1997, February 23, June 15, and July 15, 1999; and by
separate application dated October 22, 1997, as supplemented February
23, June 28, and July 15, 1999.
Brief description of amendment: This amendment implements: (1)
voltage-based repair criteria for BVPS-2 steam generator tubes similar
to the changes approved for BVPS-1 in License Amendment No. 198. The
changes revise BVPS-2 technical specifications (TSs) 4.4.5 and 3.4.6.2
and associated Bases to reflect the guidance provided in the Nuclear
Regulatory Commission's (NRC) Generic Letter 95-05, ``Voltage-Based
Repair Criteria for Westinghouse Steam Generator Tubes Affected by
Outside Diameter Stress Corrosion Cracking,'' (GL 95-05). Additionally,
BVPS-2 TS Table 4.4-2 is revised to reference TS 6.6 for reporting
requirements. (2) reduced reactor coolant system (RCS) specific
activity limits in accordance with the NRC's guidance provided in GL
95-05. The definition of Dose Equivalent I-131 is replaced with the
Improved Standard TS definition in the first sentence, and an equation
is added based on dose conversion derived from the International
Commission on Radiation Protection (ICRP) ICRP-30. TS 3.4.8, Specific
Activity, is revised by reducing the Dose Equivalent I-131 limit from
1.0 [micro] Ci [curies]/gram to 0.35 [micro] Ci [curies]/gram for the
48-hour limit and from 60 [micro] Ci [curies]/gram to 21 [micro] Ci
[curies]/gram for the maximum instantaneous limit. Item 4.a in TS Table
4.4-12, Primary Coolant Specific Activity Sample and Analysis Program;
TS Figure 3.4-1, and the Bases for TS 3/4.4.8 are also modified to
reflect the reduced Dose Equivalent I-131 limit.
The February 23, 1999, letter provided a revised control room dose
calculation in support of both the June 18, 1996, and October 22, 1997,
amendment requests. Importantly, this calculation assumed the lower
allowable primary-to-secondary leak rate limit associated with the June
18, 1996, submittal, and the reduced RCS specific activity limits
associated with the October 22, 1997, submittal. Because of this
interdependence, the changes of the first amendment request must be
implemented concurrently with those of the second in order for the
supporting analysis to remain valid. Hence, both of these license
amendment requests have been combined into this single amendment.
Date of issuance: August 18, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No: 101.
Facility Operating License No. NPF-73. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64109) and March 25, 1998 (63 FR 14485). The December 12, 1997,
February 23, June 15, June 28, and July 15, 1999, letters provided
additional information but did not change the initial proposed no
significant hazards consideration determinations or expand the
amendment requests beyond the scope of the Federal Register notices.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 18, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Date of amendment request: November 24, 1998, as supplemented by
letters dated February 25 and July 14, 1999.
Brief description of amendments: The amendments revise the
administrative sections of the Technical Specifications to reflect the
approved consolidated quality assurance program, clarify the
responsibilities of the shift technical advisor position on shift,
simplify the contents of the monthly operating report description,
complete the relocation of the fire protection requirements from
[[Page 48872]]
the Technical Specifications, and replace selected position titles with
descriptions of functional responsibility.
Date of issuance: August 26, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 198 and 209.
Facility Operating License Nos. DPR-51 and NPF-6: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4156).
The February 25 and July 14, 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 August 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of application for amendment: November 22, 1998.
Brief description of amendment: This amendment revises the reactor
thermal margin safety limit lines and flow rates stated in the St.
Lucie, Unit 1, technical specifications (TS). The amendment also
updates the reference for dose conversion factors used in Dose
Equivalent Iodine-131 calculations, makes administrative changes to the
criticality analysis uncertainty described in TS 5.6.1.a.1, updates the
analytical methods used in determining core operating limits listed in
TS 6.9.1.11, and revises the TS Bases for the steam generator pressure-
low trip setpoint.
Date of Issuance: August 18, 1999.
Effective Date: August 18, 1999.
Amendment No.: 163.
Facility Operating License No. NPF-16: Amendment revised the TS.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6696).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 18, 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.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: October 30, 1998, as
supplemented December 31, 1998, and May 12, 1999.
Brief description of amendment: The amendment approves changes to
the Improved Technical Specifications to reflect the use of Topical
Report BAW-2421 for fluence determination and changes to the low
temperature over-pressure protection limits. Changes to the CR-3
Pressure/Temperature Limits Report to reflect plant operation to 32
Effective Full Power Years were included in the submittal.
Date of issuance: August 12, 1999.
Effective date: As of date of issuance, to be implemented prior to
commencing Cycle 12 operation.
Amendment No.: 183.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71965). The supplemental letters dated December 31, 1998, and May
12, 1999, did not change the original proposed no significant hazards
consideration determination, or expand the scope of the amendment
request as originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: November 30, 1998.
Brief description of amendment: The Amendment revises Technical
Specifications (TS) to allow both doors of the containment personnel
air lock to be open during fuel movement and adds a provision for an
outage equipment hatch.
Date of issuance: August 16, 1999.
Effective date: August 16, 1999.
Amendment No.: 184.
Facility Operating License No. DPR-31: Amendment revised the TS.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4157).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: July 30, 1998, as supplemented
April 8 and July 8, 1999.
Brief description of amendment: Revises Technical Specifications
for the Control Room Emergency Ventilation System and the Ventilation
Filter Test Program.
Date of issuance: August 23, 1999.
Effective date: August 23, 1999.
Amendment No.: 185.
Facility Operating License No. DPR-31: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64115). The April 8 and July 8, 1999, supplements did not change the
original proposed no significant hazards determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: December 3, 1998, as
supplemented by letters dated March 26, April 16, May 7, May 21, June
4, June 15, and June 29, 1999.
Brief description of amendment: The amendment revises the Technical
Specification Figure 2.1-1 ``Core Protection Safety Limit,'' and Figure
2.1-3 ``Core Protection Safety Bases'' to reflect a decrease in reactor
coolant system flow resulting from a revised analysis to allow
operation of the TMI-1 facility with an average of 20 percent of the
steam generator tubes plugged, and no more than 25 percent plugged in
either generator.
Date of issuance: August 19, 1999.
Effective date: As of the date of demonstration of a satisfactory
emergency feedwater pump flow test, as described in the license
amendment and documented by the licensee, to be
[[Page 48873]]
performed during the 13R refueling outage scheduled to begin September
10, 1999, and shall be implemented within 30 days of that date.
Amendment No.: 214.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71967). The supplements dated March 26, April 16, May 7, May 21,
June 4, June 15, and June 29, 1999, are within the scope of the
original notice and do not change the proposed no significant hazards
consideration finding.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 19, 1999.
No significant hazards consideration comments received: No.
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.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: February 2, 1999 as supplemented
July 29, 1999.
Brief description of amendment: The amendment expands the scope of
systems and test requirements for post-accident reactor building sump
recirculation engineered safeguards features systems and increases the
maximum allowable leakage of TS 4.5.4 from 0.6 gallons per hour (gph)
to 15.0 gph.
Date of issuance: August 24, 1999.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 215.
Facility Operating License No. DPR-50. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14283).
The supplemental letter did not change the initial no significant
hazards consideration determination or the Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 24, 1999.
No significant hazards consideration comments received: No.
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.
Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London
County, Connecticut
Date of application for amendment: March 5, 1999.
Brief description of amendment: The amendments relocate certain
Technical Specifications (TSs) Section 6.0 administrative controls to
the NRC-approved Northeast Utilities Quality Assurance Program (NUQAP)
Topical Report. Specifically, Sections 6.2.3 (Unit 3 only), 6.5, 6.6
(partial), 6.7 (partial), and 6.10. The amendments also delete parts of
Section 6.6 and 6.7 because their requirements are duplicated in
existing regulations or elsewhere in the TSs. In addition, the
amendments modify the table of contents and other TS sections to
incorporate the aforementioned changes (e.g., correct references).
Date of issuance: August 13, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 239 and 173.
Facility Operating License Nos. DPR-65 and NPF-49: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17027).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 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.
PECO Energy Company, Public Service Electric and Gas Company Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: February 12, 1999, as
supplemented July 8, 1999. The July 8, 1999, letter provided clarifying
information and did not change the original no significant hazards
consideration determination.
Brief description of amendments: Administrative changes to correct
typographical and editorial errors in Technical Specifications
introduced in previous amendments.
Date of issuance: August 23, 1999.
Effective date: This license amendment is effective as of its date
of issuance. The amendment will be implemented within 30 days.
Amendments Nos.: 228 and 231.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24200).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
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 amendments: November 20, 1998, as
supplemented by letter dated June 25, 1998.
Brief description of amendments: These amendments modified
technical specification surveillance requirement, 3.8.1.4, to allow
increases in the minimum fuel oil required to be stored in the day
tanks for emergency diesel generators.
Date of issuance: August 23, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 185 and 159.
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
4159).
The supplemental letter provided clarifying information and did not
change the initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-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: December 19, 1997, as
supplemented June 1, 1998, and May 13, 1999.
[[Page 48874]]
Brief description of amendments: The amendments revise TS 3.4.9,
Pressurizer, to reduce the allowable pressurizer water volume for
pressurizer operability. The allowable water volume is also revised to
a percent pressurizer level of 57 percent.
Date of issuance: August 19, 1999.
Effective date: August 19, 1999, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--155; Unit 3--146.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 25, 1998 (63 FR
14488).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 19, 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.
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: September 4, 1998, as
supplemented December 8, 1998, and February 16, 1999 (PCN 493).
Brief description of amendments: The amendments revise Technical
Specification 3.4.10, Pressurizer Safety Valves, to increase the as-
found pressurizer safety valve setpoint tolerances.
Date of issuance: August 19, 1999.
Effective date: August 19, 1999, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--156; Unit 3--147.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6711). The licensee's letters dated December 8, 1998, and February
16, 1999, provided clarifications and additional information that were
within the scope of the original Federal Register notice and did not
change the staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 19, 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.
Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M.
Farley Nuclear Plant, Unit 1, Houston County, Alabama.
Date of amendment request: April 23, 1999, as supplemented by
letters dated July 22, July 30 and August 12, 1999.
Brief Description of amendment: The amendment adds an additional
condition to the license which allows Southern Nuclear Operating
Company to operate Unit 1 for Cycle 16 based on a risk-informed
approach to evaluate steam generator tube structural integrity.
Date of issuance: August 17, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 143.
Facility Operating License No. NPF-2: Amendment revises the
Facility Operating License to add a license condition.
Date of initial notice in Federal Register: June 16, 1999 (64 FR
32291).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
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: March 22, 1999, as supplemented July 15,
1999.
Brief description of amendments: The amendments revised Technical
Specification 3/4.7.1.6, ``Atmospheric Steam Relief Valves,'' and added
a new Technical Specification for atmospheric steam relief valve
instrumentation, to ensure that the automatic feature of the steam
generator power-operated relief valves (i.e., the atmospheric steam
relief valves) remains operable during Modes 1 and 2.
Date of issuance: August 19, 1999.
Effective date: August 19, 1999, to be implemented within 30 days.
Amendment Nos.: Unit 1--114; Unit 2--102.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 21, 1999 (64 FR
19565).
The July 15, 1999, supplement provided revised Technical
Specification pages and clarifying information that was within the
scope of the original Federal Register notice and did not change the
staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendment: June 3, 1999 (TS 397).
Brief description of amendment: The Amendments change the Technical
Specifications (TS) by reducing the Allowable Value used for Reactor
Vessel Water Level--Low, Level 3 for several instrument functions.
Date of issuance: August 16, 1999.
Effective date: August 16, 1999.
Amendment Nos.: 260 and 219.
Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revise the TS.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38037).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: April 16, 1999, as supplemented
June 9, 1999.
Brief description of amendment: The amendment clarifies the
inservice inspection requirements regarding the granting of relief from
the American Society of Mechanical Engineers (ASME) Code requirements
by the NRC. The amendment also made changes to reflect previous NRC
approval of the use of ASME Code Case N-560.
Date of Issuance: August 13, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
[[Page 48875]]
Amendment No.: 172.
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38037).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated August 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: June 24, 1999.
Brief description of amendment: The amendment clarifies the basis
for the reactor protection system bypass of the turbine stop valve
closure and turbine control valve fast closure scram signals at low
power. The amendment clarifies that the analytical basis for this
bypass corresponds to a fraction of reactor rated thermal power and not
other measures of power, for instance, turbine power.
Date of Issuance: August 13, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 173.
Facility Operating License No. DPR-28.: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38038).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated August 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Dated at Rockville, Maryland, this 1st day of September 1999.
For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 99-23300 Filed 9-7-99; 8:45 am]
BILLING CODE 7590-01-P