[Federal Register Volume 59, Number 158 (Wednesday, August 17, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10817]
[[Page Unknown]]
[Federal Register: August 17, 1994]
_______________________________________________________________________
Part II
Nuclear Regulatory Commission
_______________________________________________________________________
Operating Licenses, Amendments; No Significant Hazards Considerations;
Biweekly Notices
UNITED STATES NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
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 July 25, 1994, through August 5, 1994. The
last biweekly notice was published on August 3, 1994.
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By September 16, 1994, 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 20555 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, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 20555, and at the local public document
room for the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: May 4, 1994
Description of amendment requests: These amendment requests would
revise Limiting Condition for Operation (LCO) 3.4.8.3 and Surveillance
Requirement 4.4.8.3.1, ``Overpressure Protection Systems.''
Specifically, the LCO and surveillance requirements are revised to
clarify that both shutdown cooling system (SCS) suction line relief
valves shall be OPERABLE and aligned to provide overpressure protection
not only during reactor (RCS) cooldown or heatup evolutions, but also
during any steady state temperature periods maintained in the course of
RCS cooldown or heatup evolutions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1 -- Involve a significant increase in the probability
or consequence of an accident previously evaluated.
The proposed amendments provide further clarification of the
Technical Specifications and represent an additional operating
limitation. Incorporating the noted clarification will not change
the bases or assumptions contained in the safety analysis for this
system. The most limiting low-temperature overpressure protection
(LTOP) transients, the starting of an idle reactor coolant pump
(RCP) and the inadvertent actuation of two high pressure safety
injection (HPSI) pumps into a solid RCS, are not affected by the
proposed clarification. Therefore, the proposed amendments do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Standard 2 -- Create the possibility of a new or different kind
of accident from any accident previously evaluated.
Clarifying the applicability of the LCO's and surveillance for
steady state periods achieved and maintained during either a heatup
or cooldown evolution does not modify the design or operation of
plant equipment. No new or different failure modes will be
introduced by incorporating this clarification into the LCO and
surveillance requirement. Therefore, the proposed amendments will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Standard 3 -- Involve a significant reduction in a margin of
safety.
The clarification will enhance LCO 3.4.8.3 and Surveillance
Requirement 4.4.8.3.1 for heatup and cooldown evolutions by ensuring
operators are aware of this applicability during periods of steady
state conditions. This clarification does not involve a change to
safety limits, setpoints, or design margins. As such, the proposed
amendments will not involve a significant reduction in a margin of
safety at PVNGS.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: June 17, 1994
Description of amendment requests: The proposed amendments would
increase the minimum nitrogen accumulator pressure for the atmospheric
dump valves (ADVs), as stated in the surveillance requirements of
Technical Specification (TS) 3/4.7.1.6. The change to the Bases
increases the minimum time the ADV accumulators must be operable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1 -- Involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed Technical Specification change in the nitrogen
accumulator supply minimum pressure will not increase the
probability or consequences of any accident previously analyzed. The
nitrogen accumulator pressure is normally maintained between 650-680
psig. Nitrogen pressure from the accumulator is reduced to 105 psig
prior to use in the operation of the ADVs. The pressure reduction
will remain the same with the higher minimum accumulator pressure.
Standard 2 -- Create the possibility of a new or different kind
of accident from any accident previously evaluated.
Increasing the nitrogen accumulator minimum pressure does not
create any new or different accidents than those previously
evaluated. The normal air supply (the Instrument Air System) to the
ADV is maintained between 105 to 125 psig. Currently, nitrogen from
the accumulator is reduced to 105 psig prior to use in the ADV. The
increased minimum pressure in the accumulator will still be reduced
to 105 psig prior to use in the ADV.
Standard 3 -- Involve a significant reduction in a margin of
safety.
The limitation on maintaining the nitrogen accumulator at a
certain pressure is to ensure that a sufficient volume of nitrogen
is in the accumulator to operate the associated ADV. Maintaining a
higher minimum pressure ensures that sufficient nitrogen will be
available to maintain the unit at HOT STANDBY for four hours and an
additional 9.3 hours to reach COLD SHUTDOWN under natural
circulation conditions in the event of failure of the normal control
air system. Therefore, the proposed change does not involve a
reduction in a margin of safety.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: July 12, 1994
Description of amendment requests: The proposed amendment would
enhance the PVNGS Technical Specifications (TS) by adding a limiting
condition for operation (LCO) action statement to Entry VIII B of Table
3.3-3, ``Engineered Safety Features Actuation System Instrumentation.''
The proposed action statement would enhance safe plant operation by
requiring timely plant shutdown if more than one of the new solid state
degraded voltage relays in either train of 4.16kV are inoperable or not
energized.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Involve a significant increase in the probability or
consequence of an accident previously evaluated:
The proposed amendment will add an action statement to TS Table
3.3-3 entry VIII B which would allow eight hours to effect repairs.
This action statement would be entered if more than one of the
required four degraded voltage relays on either 4.16 kv bus is
inoperable or not energized. If the eight hour allowed outage time
is not met, the unit is placed in Hot Standby within six hours and
in Cold Shutdown within the next thirty hours. Technical
Specification 3.8.3.1 currently allows eight hours to restore a 4.16
kv bus in the event of a loss of power to that bus. The loss of
degraded voltage relays on that bus does not impact plant nuclear
safety any more than the loss of the bus itself. Furthermore, even
with the loss of all four degraded voltage relays monitoring one
4.16 kv bus (for example, due to a blown 125 vdc circuit fuse), the
loss-of-voltage relays on that bus, and the degraded voltage relays,
as well as the loss-of-voltage relays monitoring the other bus would
be unaffected. None of the UFSAR chapter 15 accident analyses are
affected by this proposed amendment. The existing TS requirements
and those components to which they apply are not altered by this TS
amendment. There are no changes to the maintenance, surveillance,
and/or qualification of any component/function in Table 3.3-3.
Therefore, the addition of this proposed eight hour action statement
to Table 3.3-3 entry VIII B does not increase the probability of
occurrence or the consequences of any previously evaluated accident.
Standard 2--Create the possibility of a new or different kind of
accident from any accident previously evaluated:
The TS requirements and the components to which they apply are
not altered by this amendment. The new solid state degraded voltage
relays in each 4.16 kv bus were installed under the 10 CFR 50.59
change process. APS [Arizona Public Service Company] determined that
the installation created no unreviewed safety question. This
amendment has no impact on plant maintenance, testing, shutdown
equipment, or component qualification. Plant operational safety is
enhanced by this amendment. Therefore, the possibility of a new or
different kind of accident is not created by this amendment.
Standard 3--Involve a significant reduction in a margin of
safety:
The TS does not alter existing TS requirements or those
components to which they apply. More specifically, there is no
impact on safe plant shutdown, maintenance, containment isolation
capability, containment leakage rate, or the operability of safety
related valves. Therefore, the addition of the proposed action
statement to the TS will not involve reduction in a margin of safety
for fission product release to the atmosphere.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Basis for proposed no significant hazards consideration
determination:
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: June 8, 1994
Description of amendments request: The proposed amendment would
revise the Calvert Cliffs Nuclear Power Plant (CCNPP) Units 1 and 2
Technical Specification (TS) Section 4.7.1.2.c to extend the interval
for three Auxiliary Feedwater (AFW) surveillance requirements from 18
to 24 months. Specifically, TS Section 4.7.2.c.1 requires the
verification of each automatic valve in the flowpath actuate to its
correct position and each AFW pump automatically start upon receipt of
each AFW actuation system test signal; TS Section 4.7.2.c.2 requires
verification that the AFW system is capable of providing a minimum 300
gallons per minute nominal flow to each leg. This request is one of a
series of proposed license amendments that would eliminate the need for
mid-cycle surveillance outages by extending 18-month frequency
surveillances to every refueling outage (nominally each 24 months).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The Auxiliary Feedwater (AFW) System provides a safety-related
source of feedwater to the steam generators to mitigate design basis
accidents involving loss of Main Feedwater. Failure of the AFW
System is not an initiator for any previously analyzed accident.
Therefore, the proposed change does not involve an increase in the
probability of an accident previously evaluated.
A historical review of surveillance test results and system
performance indicates that the AFW System is very reliable. In
addition, monthly surveillances of the AFW System will continue to
verify proper pump and valve operation. The AFW System reliability
and monthly surveillances provide assurance that undetected system
degradation will not occur between 24-month surveillances.
Therefore, the AFW System will continue to perform its safety
function and there will be no significant increase in the
consequences of accidents. Therefore, the proposed Technical
Specification changes do not increase the probability or
consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated?
This requested revision to increase the interval for some AFW
surveillances from 18 to 24 months does not involve a significant
change in the design or operation of the plant. No hardware is being
added to the plant as part of the proposed change. The proposed
change will not introduce any new accident initiators. Therefore,
this change would not create the possibility of a new or different
type of accident from any accident previously evaluated.
3. Does operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
The AFW System provides a margin of safety by providing a
safety-related alternate supply of feedwater to the steam generator
for removal of decay heat and cooldown of the Reactor Coolant
System. The proposed changes do not affect the operation or design
of the AFW System. Monthly surveillances and historical data provide
assurance that the reduction in surveillance frequency will not
adversely affect our ability to detect degradation in the system.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael L. Boyle
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: April 29, 1994
Description of amendment request: This amendment is an additional
followup to the amendment request of May 29, 1992, published in the
Federal Register on July 8, 1992 (57FR30242), which changed the
Technical Specifications Section 1.0, Definitions, to accommodate a 24-
month fuel cycle and which proposed the extension of the test intervals
for specific surveillance tests. This amendment proposes extending the
surveillance intervals to 24 months for the following additional
surveillance tests:(1) Calibrate and test channels for Auxiliary
Feedwater (AFW) initiation on steam generator water level (low-low).(2)
Test channels for Auxiliary Feedwater initiation on trip of main
feedwater pumps.The licensee's amendment proposal of November 25, 1992,
requested approval for extending the surveillance interval of the
Auxiliary Feedwater System to accommodate a 24-month fuel cycle and the
approved change was issued in License Amendment No. 166. Subsequently,
the licensee determined that two additional surveillances associated
with this system had not been identified in the November 25, 1992,
request. This amendment proposal requests approval of the additional
surveillances. The changes requested by the licensee are in accordance
with Generic Letter 91-04, ``Changes in Technical Specification
Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''
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 proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
The test results over the last four refuelings confirmed system
operability with only one failure. This failure would not have
impaired the ability of the auxiliary feedwater system to perform
its intended safety function. The auxiliary feedwater system is
redundant and diverse. The failure in the turbine driven pump did
not impact the motor driven pumps.
Based on the historical test data, it is concluded that no
significant increase in the probability or consequences of an
accident would be incurred by extending the operating cycle due to
an increased surveillance interval.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The failure noted from the past test data appears random in
nature and would not have defeated the redundancy in design that
exists in the AFW system. The AFW system would have been capable of
performing its intended safety function and therefore a new or
different kind of accident would not have been created.
3. There has been no reduction in the margin of safety.
Past historical data demonstrates that the AFW systems would
perform their safety function for an extended operating cycle should
the surveillance period be extended by several months.
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.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Pao Tsin Kuo
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: June 16, 1994
Description of amendment request: The proposed amendment would
revise License Condition 2.K of the license issued August 24, 1981, to
provide for compliance with the NRC-approved fire protection program as
described in the Updated Final Safety Analysis Report and for making
changes to the NRC-approved fire protection program; would delete fire
protection Technical Specification (TS) Sections 3.13 and 4.14 which
contain limiting conditions for operation and surveillance
requirements, respectively, for the high-pressure water fire protection
system, fire protection spray systems, penetration fire barriers, fire
detection systems, fire hose stations and hydrants, and the cable
spreading room halon system; would delete Section 6.2.2.f which
contains fire brigade staffing requirements; would delete Section 6.4.2
which contains fire brigade training requirements; would add Section
6.5.1.6.1 to add fire protection program responsibilities to the
Station Nuclear Committee; would add Section 6.8.1.e to require written
procedures and administrative policies for the fire protection program;
would delete Section 6.9.2.b which requires a Special Report for
inoperable fire protection and detection equipment; and would make
corresponding changes to the Table of Contents and List of Tables.
Generic Letter (GL) 86-10, dated April 24, 1986, and GL 88-12,
dated August 2, 1988, from the NRC provided guidance to licensees to
request removal of the fire protection TS. The licensee's proposed
amendment is in response to these GLs.
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 Commission has provided guidance concerning the application
of the standards for determining whether a ``Significant Hazards
Consideration'' exists by providing certain examples in 51 FR 7744
(dated March 6, 1986). Example (vii) of those involving no
significant hazards considerations relates to ``a change to conform
a license to changes in the regulations, where the license change
results in very minor changes to facility operations clearly in
keeping with the regulations.''
In this case, NRC Generic Letters 86-10 and 88-12, although not
regulations, provide pertinent guidance relative to the above
described proposed changes and implementation of the NRC fire
protection regulations of 10 CFR 50.48(a). Specifically, the generic
letters allow licensees to delete fire protection related technical
specifications, provided that administrative requirements are added
to technical specifications and a license condition is provided that
requires the implementation and maintenance in effect of the
approved fire protection program. Further, the generic letters
provide for inclusion of the fire protection program into the UFSAR
[Updated Final Safety Analysis Report] and permits future changes to
the fire protection program without prior NRC approval, all as
provided by the license condition and in accordance with the
provisions of 10 CFR 50.59. Therefore, since the actions required by
the generic letters have been taken and conform the license to the
current interpretation of NRC fire protection regulations as
described in Generic Letters 86-10 and 88-12, with no changes to
facility operations, these proposed changes are in accordance with
Example (vii) above.
In accordance with the requirements of 10 CFR 50.92, the
proposed changes to the Technical Specifications are deemed not to
involve any ``Significant Hazards Considerations'' because operation
of Indian Point Unit No. 2 in accordance with these changes would
not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The fire protection program requirements are not affected in
that the function, operation or surveillance requirements for any
fire protection system or component are not being altered. The
proposed changes simply relocate these requirements from the
Technical Specifications to the UFSAR, are administrative in nature,
and do not affect any other current plant equipment or practices.
Therefore, the conclusions of current accident analyses are not
affected. Further, as permitted by the proposed License Condition
2.K, changes in the NRC-approved fire protection program will
require an evaluation per the criteria of 10 CFR 50.59 to determine
that the proposed change will not involve an unreviewed safety
question. Therefore, future changes to the fire protection program
will be evaluated in accordance with appropriate criteria.
(2) Create the possibility for a new or different kind of
accident from any previously evaluated.
The proposed changes introduce no new mode of plant operation,
do not involve physical modification to any structure, system or
component, do not affect the function, operation or surveillance
requirements for any equipment necessary for safe operation or
shutdown of the plant or of fire protection equipment which protects
such equipment, and do not involve any changes to setpoints or
operating parameters. The changes are administrative only and all
existing fire protection requirements are maintained. Therefore, the
changes can not result in an unanalyzed accident. Further, as
permitted by the proposed License Condition 2.K, changes in the NRC-
approved fire protection program will require an evaluation per the
criteria of 10 CFR 50.59 to determine that the proposed change will
not involve an unreviewed safety question. Therefore, future changes
to the fire protection program will be evaluated in accordance with
appropriate criteria.
(3) Involve a significant reduction in the margin of safety.
The existing fire protection program operability and
surveillance requirements are retained as they are contained in the
FPPP [Fire Protection Program Plan], and compliance will continue
through proposed License Condition 2.K. Therefore, no margins of
safety established by design or verified by testing to ensure
operability of fire protection systems or components are affected.
Further, as permitted by the proposed License Condition 2.K, changes
in the NRC-approved fire protection program will require an
evaluation per the criteria of 10 CFR 50.59 to determine that the
proposed change will not involve an unreviewed safety question.
Therefore, future changes to the fire protection program will be
evaluated in accordance with appropriate criteria.
Based on the above discussion, since these proposed changes to
the Indian Point Unit No. 2 Technical Specifications satisfy the
criteria specified in 10 CFR 50.92, are similar to an example
provided by the Commission of a change which involves ``No
Significant Hazards Considerations'', and are not similar to any
examples that involve a ``Significant Hazards Consideration'', Con
Edison has determined that this amendment application does not
involve any ``Significant Hazards Considerations.''
The proposed Technical Specification changes have been reviewed
by the Station Nuclear Safety Committee and the Con Edison Nuclear
Facilities Safety Committee. Both committees concur that these
proposed changes do not represent any ``Significant Hazards
Considerations.''
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.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Michael L. Boyle
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: July 8, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification Section 3.7, Auxiliary Electrical
Systems to clarify offsite power availability requirements and to
revise emergency diesel generator fuel oil availability requirements.
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:
Consistent with the requirements of 10 CFR 50.92, the enclosed
application involves no significant hazards based on the following
information:
1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
Neither the probability nor the consequence of an accident
previously analyzed is increased due to the proposed changes. There
are no changes on the existing offsite power supply configuration or
on the existing diesel fuel oil supply system or inventory
requirements. This proposed amendment will allow for three diesel
operation when a fuel oil storage tank or transfer pump is
unavailable. In the event of an accident at this time, the three
diesel operation would allow for more than minimum safeguards to be
available, with maximum safeguards available for the first part of
the event.
2) Does the proposed license amendment create the possibility of
a new or different kind of accident from any previously evaluated?
Response:
The existing 138 kV and 13.8 kV offsite power reliability is
maintained with this change. There is no impact on availability of
the alternate AC system, the three gas turbines, with this change.
This change is consistent with the original licensing basis that the
AEC accepted for the diesel fuel oil supply system.
3) Does the proposed amendment involve a significant reduction
in the margin of safety?
Response:
The proposed amendment does not involve a significant reduction
in the margin of safety. The proposed amendment maintains the
reliability of the preferred 138 kV and 13.8 kV offsite power and is
consistent with the original licensing basis for diesel fuel oil
inventory.
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.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Robert A. Capra
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of amendment request: July 29, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 3/4.4.5 and 3/4.4.6.2 and
associated bases to allow the implementation of interim steam generator
tube plugging criteria for the tube support elevations during cycle 11.
The allowed primary-to-secondary operational leakage from any one steam
generator is proposed to be reduced from 500 gallons per day (gpd) to
150 gpd. The total allowed primary-to-secondary operational leakage
from all steam generators would be reduced from one gallon per minute
(1440 gpd) to 450 gpd.
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?
Testing of model boiler specimens for free span tubing (no tube
support plate [TSP] restraint) at room temperature conditions
show[s] burst pressures in excess of 5000 psi for indications of
outer diameter stress corrosion cracking with voltage measurement as
high as 19 volts. Burst testing performed on intersections pulled
from BVPS [Beaver Valley Power Station] with up to a 2.7 volt
indication shows measured burst pressure in excess of 6600 psi at
room temperature. Burst testing performed on pulled tubes from other
plants with up to 7.5 volt indications show[s] burst pressures in
excess of 6300 psi at room temperatures. Correcting for the effects
of temperature on material properties and minimum strength levels
(as the burst testing was done at room temperature), tube burst
capability significantly exceeds the safety factor requirements of
RG [Regulatory Guide] 1.121. As stated earlier, tube burst criteria
are inherently satisfied during normal operating conditions due to
the proximity of the TSP. Test data indicates that tube burst cannot
occur within the TSP, even for tubes which have 100 percent through
wall electric discharge machining (EDM) notches, 0.75 inch long,
provided that the TSP is adjacent to the notched area. Since tube to
TSP proximity precludes tube burst during normal operating
conditions, use of the criteria must retain tube integrity
characteristics which maintain a margin of safety of 1.43 times the
bounding faulted condition steam line break (SLB) pressure
differential. As previously stated, the RG 1.121 criterion requiring
maintenance of a safety factor of 1.43 times the SLB pressure
differential on tube burst is satisfied by 7/8 inch diameter tubing
with bobbin coil indications with signal amplitudes less than 8.82
volts regardless of the indicated depth measurement. The plugging
criteria (resulting in a projected end-of-cycle [EOC] voltage)
compares favorably with the 8.82 volt structural limit considering
the extremely slow apparent voltage growth rate of indications at
BVPS. Using the established methodology of RG 1.121, the structural
limit is reduced by allowances for uncertainty and growth to develop
a beginning-of-cycle (BOC) repair limit which should preclude
indications at EOC conditions which exceed the structural limit. The
non-destructive examination (NDE) uncertainty component is 20.5
percent and is based on the EPRI [Electric Power Research Institute]
Alternate Repair Criteria (ARC). A bounding growth allowance of 40
percent will be applied. This value is conservative for BVPS Unit 1.
The BOC maximum allowable repair limit should not permit the
existence of EOC indications (when the 40 percent growth and 20.5
percent uncertainty allowances are applied) which exceed the 8.82
volt structural limit. By adding NDE uncertainty allowances and an
allowance for crack growth to the repair limit, the structural limit
can be validated. Therefore, the maximum allowable BOC repair limit
(RL) based on the structural limit of 8.82 volts can be represented
by the expression:
RL + (0.205 X RL) + (0.40 x RL) = 8.82 volts, or the maximum
allowable BOC repair limit can be expressed as:
RL = 8.82 volt structural limit/1.605 = 5.5 volts.
It is reasonable that this repair limit (5.5 volts) could be
applied for IPC [interim plugging criterion] implementation to
repair bobbin indications greater than 1.0 or 2.0 volts independent
of RPC [rotating pancake coil] confirmation of the indication. The
analyses were performed based on a 1.0 or 2.0 volt repair limit.
Duquesne Light Company has chosen to use a steam generator tube
repair limit of 1.0 volt. Conservatively, an upper limit of 3.6
volts will be used to assess tube integrity for those bobbin
indications which are above 1.0 volt but do not have confirming RPC
calls. This 3.6 volt upper limit for non-confirmed RPC calls is
consistent with other recently approved IPC programs for the two
other plants with 7/8 inch tubing that currently implement IPCs.
Since the upper bound for repair of non-confirmed RPC is limited to
a value far less than the limit associated with a full alternate
criteria, the establishment of the repair limits are [is] judged to
be independent of the pulled tube data base used.
The conservatism of the growth allowance used to develop the
repair limit is shown by the most recent BVPS eddy current data. The
average voltage growth for all indications was 16 percent while the
average voltage growth for indications greater than 0.75 volts at
BOC was 6 percent. The largest overall voltage growth in a
particular steam generator was found in the ``A'' steam generator,
which had an overall average growth of 25 percent. Only two tubes
had an absolute voltage growth which exceeded 1.0 volt for Cycle 9.
The maximum absolute voltage growth in the 1993 inspection was
recorded to be 1.18 volts. Each of the last three inspections, which
included 100 percent of all hot leg tubes, showed decreasing voltage
growth trends in each successive inspection for all categories;
overall voltage growth, growth of BOC indications less than 0.75
volts, and growth of indications greater than 0.75 volts. The
decreasing voltage growth rate trend data indicates that DLC has
good control of the ODSCC [outer diameter stress corrosion cracking]
occurring in the BVPS Unit 1 steam generators and also implies that
atypical voltage growth of a few indications is unlikely.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated main
SLB [steam line break] outside of containment but upstream of the
main steam isolation valve (MSIV) represents the most limiting
radiological condition relative to the IPC. In support of
implementation of the interim plugging criteria, it will be
determined whether the distribution of cracking indications at the
TSP intersections at the end of Cycle 11 are [is] projected to be
such that primary-to-secondary leakage would result in site boundary
doses within a small fraction of the 10 CFR 100 guidelines. A
separate calculation has determined this allowable SLB leakage limit
to be 6.6 gpm in the faulted loop. This limit was calculated using
the Technical Specification RCS [reactor coolant system] Iodine-131
activity level of 1.0 micro Curies per gram dose equivalent Iodine-
131 and the recommended Iodine-131 transient spiking values
consistent with NUREG-0800. The projected SLB leakage rate
calculation methodology prescribed in Section 3.3 of draft NUREG-
1477 will be used to calculate EOC leakage. The log-logistic
probability of leakage correlation will be used to establish the SLB
leak rate used for comparison with the 6.6 gpm faulted loop
allowable limit. Due to the relatively low voltage levels of
indications at BVPS and low voltage growth rates, it is expected
that the actual calculated leakage values will be far less than this
limit. Additionally, the current Iodine-131 levels as of May 1994 at
BVPS are about 1000 times less than the Technical Specification
limit of 1.0.
Application of the criteria requires the projection of
postulated SLB leakage, based on the projected EOC voltage
distribution for the upcoming cycle. Projected EOC voltage
distribution is developed using the most recent EOC eddy current
results and a voltage measurement uncertainty. Data indicate that a
threshold voltage of 2.8 volts would result in through wall cracks
long enough to leak at steam line break conditions. Draft NUREG-1477
requires that all indications to which the IPC are applied must be
included in the leakage projection. Tube pull results from another
plant with 7/8 inch tubing with a substantial voltage growth data
base have shown that tube wall degradation of greater than 40
percent through wall was readily detectable either by the bobbin or
RPC probe. The tube with maximum through wall penetration of 56
percent (42 percent average) had a voltage of 2.02 volts. This
indication also was the largest recorded bobbin voltage from the EOC
eddy current data. Based on the BVPS pulled tube and industry pulled
tube data supporting a lower threshold for SLB leakage of 2.8 volts,
inclusion of all IPC intersections in the leakage model is quite
conservative. The ODSCC occurring at BVPS has historically resulted
in relatively low voltage levels and has exhibited decreasing
voltage growth trends over the last three inspections. BVPS has not
identified ODSCC as a contributor to operational leakage. The
current leakage levels at BVPS are negligible (less than 1 gpd). In
order to satisfy the requirements of draft NUREG-1477, EOC 10 eddy
current data will be used to calculate the projected SLB leakage
according to draft NUREG-1477 methodology. Leakage calculated using
the recommended EPRI leakage correlation will also be provided.
Duquesne Light Company is requesting that the NRC review and approve
the EPRI SLB leakage calculation methodology. Sufficient
justification is included to establish acceptability of the EPRI
leakage correlation based on criteria provided by the NRC in the
February 8, 1994, Industry/NRC working meeting on the voltage based
criteria.
In order to assess the sensitivity of application of the voltage
based criteria upon SLB leakage, the EOC 9 eddy current results were
used to calculate postulated EOC 10 leakage using both the NUREG-
1477 methodology and EPRI correlation assuming that a 1.0 or 2.0
volt plugging limit were implemented at the BOC 10.
Results indicate SLB leakage of 0.46 gpm and 0.044 gpm using the
NUREG and EPRI methodologies with an assumed probability of
detection (POD) of 0.6 for a 2.0 volt repair limit. Since Duquesne
Light Company has chosen to limit the voltage based plugging limit
at 1.0 volt, EOC 11 SLB leakage is analyzed to be approximately 5
percent lower than the calculated SLB leakage with a 2.0 volt repair
limit.
Therefore, implementation of the interim plugging criteria does
not adversely affect steam generator tube integrity and
implementation will be shown to result in acceptable dose
consequences, therefore, the proposed amendment does not result in
any 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?
Implementation of the proposed steam generator tube interim TSP
plugging criteria does not introduce any significant changes to the
plant design basis. Use of the criteria does not provide a mechanism
which could result in an accident outside of the region of the TSP
elevations; no ODSCC that has been identified at the TSP has been
detected outside the thickness of the TSPs. Neither a single or
multiple tube rupture event would be expected in a steam generator
in which the plugging criteria has been applied (during all plant
conditions).
Specifically, Duquesne Light Company will implement a maximum
leakage rate limit of 150 gpd per steam generator to help preclude
the potential for excessive leakage during all plant conditions. The
technical specification limits on primary-to-secondary leakage at
operating conditions are to be a maximum of 450 gpd for all steam
generators, or, a maximum of 150 gpd for any one steam generator.
The RG 1.121 criterion for establishing operational leakage rate
limits that require plant shutdown are based upon leak-before-break
considerations to detect a free span crack before potential tube
rupture during faulted plant conditions. The 150 gpd limit should
provide for leakage detection and plant shutdown in the event of the
occurrence of an unexpected single crack resulting in leakage that
is associated with the longest permissible crack length. RG 1.121
acceptance criteria for establishing operating leakage limits are
based on leak-before-break considerations such that plant shutdown
is initiated if the leakage associated with the longest permissible
crack is exceeded.
The single through wall crack lengths that result in tube burst
at 1.43 times the steam line break pressure differential and SLB
pressure differential alone are approximately 0.57 inch and 0.84
inch, respectively. A leak rate of 150 gpd will provide for
detection of 0.41 inch long cracks at nominal leak rates and 0.62
inch long cracks at the lower 95 percent confidence level leak
rates. Since tube burst is precluded during normal operation due to
the proximity of the TSP to the tube and the potential exists for
the crevice to become uncovered during SLB conditions, the leakage
from the maximum permissible crack must preclude tube burst at SLB
conditions. Thus, the 150 gpd limit provides for plant shutdown
prior to reaching critical crack lengths for SLB conditions using
the lower 95 percent leakage data. Additionally, this leak-before-
break evaluation assumes that the entire crevice area is uncovered
during blowdown. Partial uncovery will provide benefit to the burst
capacity of the intersection. Analyses have shown that only a small
percentage of the TSPs are deflected greater than the TSP thickness
during a postulated SLB.
Steam generator tube integrity continues to be maintained
through inservice inspection and primary-to-secondary leakage
monitoring, therefore, the possibility of a new or different kind of
accident from any accident previously developed is not created.
3. Does the change involve a significant reduction in a margin
of safety?
The use of the voltage based bobbin probe interim TSP elevation
plugging criteria is demonstrated to maintain steam generator tube
integrity commensurate with the requirements of RG 1.121. RG 1.121
describes a method acceptable to the NRC staff for meeting GDCs 14,
15, 31, and 32 by reducing the probability or the consequences of
steam generator tube rupture. This is accomplished by determining
the limiting conditions of degradation of steam generator tubing, as
established by inservice inspection, for which tubes with
unacceptable cracking should be removed from service. Upon
implementation of the criteria, even under the worst case
conditions, the occurrence of ODSCC at the TSP elevations is not
expected to lead to a steam generator tube rupture event during
normal or faulted plant conditions. The EOC distribution of crack
indications at the TSP elevations will be confirmed to result in
acceptable primary-to-secondary leakage during all plant conditions
and that radiological consequences are not adversely impacted.
In addressing the combined effects of loss of coolant accident
(LOCA) and safe shutdown earthquake (SSE) on the steam generator
component (as required by GDC 2), it has been determined that tube
collapse may occur in the steam generators at some plants. This is
the case as the TSP may become deformed as a result oflateral loads
at the wedge supports at the periphery of the plate due to the
combined effects of the LOCA rarefaction wave and SSE loadings.
Then, the resulting pressure differential on the deformed tubes may
cause some of the tubes to collapse.
There are two issues associated with steam generator tube
collapse. First, the collapse of steam generator tubing reduces the
RCS flow area through the tubes. The reduction in flow area
increases the resistance to flow of steam from the core during a
LOCA which, in turn, may potentially increase peak clad temperature
(PCT). Second, there is a potential that partial through wall cracks
in tubes could progress to through wall cracks during tube
deformation or collapse.
Consequently, since the leak-before-break methodology is
applicable to the BVPS reactor coolant loop piping, the probability
of breaks in the primary loop piping is sufficiently low that they
need not be considered in the structural design of the plant. The
limiting LOCA event becomes either the accumulator line break or the
pressurizer surge line break. LOCA loads for the primary pipe breaks
were used to bound the conditions at BVPS for smaller breaks. The
results of the analysis using the larger break inputs show that the
LOCA loads were found to be of insufficient magnitude to result in
steam generator tube collapse or significant deformation. The LOCA
and SSE tube collapse evaluation performed for another plant with
Series 51 steam generators using bounding input conditions (large
break loadings) is considered applicable to BVPS.
Addressing RG 1.83 considerations, implementation of the bobbin
probe voltage based interim tube plugging criteria is supplemented
by: enhanced eddy current inspection guidelines to provide
consistency in voltage normalization, a 100 percent eddy current
inspection sample size at the TSP elevations, and RPC inspection
requirements for the larger indications left inservice to
characterize the principal degradation as ODSCC.
As noted previously, implementation of the TSP elevation
plugging criteria will decrease the number of tubes which must be
repaired. The installation of steam generator tube plugs reduces the
RCS flow margin. Thus, the implementation of the alternate plugging
criteria will maintain the margin of flow that would otherwise be
reduced in the event of increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to plant safety as defined in the Final Safety
Analysis Report or any Bases of the Technical Specifications.
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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler, Director
Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One,
Unit 1, Pope County, Arkansas
Date of amendment request: June 22, 1994
Description of amendment request: The proposed amendment revises
technical specifications (TSs) related to the emergency feedwater
system (EFW). The proposed changes extend the allowable outage time
when one EFW train is inoperable from 36 hours to 72 hours and adapt
other EFW sections from the ``Restructured Standard Technical
Specifications for B&W Plants'' to the Arkansas Nuclear One, Unit 1
(ANO-1) format.
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 Emergency Feedwater (EFW) system mitigates the consequences
of any event with a loss of normal feedwater. This system is not the
initiator of any previously analyzed accident, and therefore,
changes to the specifications applicable to the EFW system present
no significant increase in the probability of any previously
evaluated accident.
The changes that revise the required Actions and Allowable
Outage Times associated with the EFW system have been evaluated for
their effect on the Core Damage Frequency (CDF) previously
calculated in the ANO-1 Probabilistic Risk Assessment (PRA). The new
ANO-1 CDF values, incorporating the proposed AOT extension, are
4.73E-05 (for the turbine-driven EFW pump) and 4.70E-05 (for the
motor-driven pump). These values do not exceed the NRC Safety Goal
of 1.0E-04 per reactor year, as stated in the Federal Register
50FR32138. The delta CDF associated with these changes (6.16E-07 for
the turbine-driven EFW pump and 3.04E-07 for the motor-driven EFW
pump) have been evaluated with respect to criteria contained in
SECY-91-270, dated August 27, 1991, and NUMARC 91-04, dated January
1992, and fall within the category of events of low risk
significance requiring no compensatory measures. This evaluation has
shown the risk associated with the proposed changes to pose no undue
risk to public health and safety, to be categorized as having low
risk significance, and involve no significant increase in the
consequences of an accident previously evaluated.
The changes revising the Limiting Conditions for Operation
result in more restrictive controls on the operability of the motor-
driven EFW pump. The previous specification required operability of
both EFW pumps when the reactor was heated above 280 deg.F. The
proposed change requires the operability of the motor-driven EFW
pump whenever the unit is above the cold shutdown condition and any
steam generator is relied upon for heat removal. With this change,
the motor-driven EFW pump is now required to be operable in a
condition not previously specified, constituting an additional
requirement not previously specified. This change does not involve a
significant increase in the consequences of an accident previously
evaluated.
The changes revising the Limiting Conditions for Operation also
incorporate an Allowable Outage Time for the turbine-driven EFW pump
steam supply valves which was not previously specified. The 7 day
AOT is reasonable based on:
1. The redundant steam supply (from the opposite steam
generator) to the turbine-driven EFW pump is operable,
2. The motor-driven EFW pump is operable, and
3. The probability of an event occurring that would require the
inoperable steam supply valve to actuate is relatively low.
The changes to the surveillance specifications clarify the
proper conditions required for the operability test of the turbine-
driven EFW pump, and revise the requirement for the verification of
proper EFW flow path valve alignment. The change clarifying the test
conditions is required to ensure a sufficient steam supply to the
turbine-driven EFW pump to perform the test. During plant startup,
from an RCS temperature of 280 deg.F to an RCS temperature of
approximately 525 deg.F (corresponding to a steam generator pressure
of approximately 830 psig) the turbine-driven EFW pump is classified
as available until operability is proven by successful completion of
the surveillance requirement. The proposed changes state that the
EFW pumps and their associated flow paths shall be operable when the
RCS is above the cold shutdown condition with any steam generator
relied upon for heat removal (motor-driven EFW pump) and when RCS
temperature is greater than or equal to 280 deg.F (turbine-driven
EFW pump). This specification requires that the flow paths be
properly aligned to maintain operability and is as restrictive as
the current TS 4.8.1.c. The revised specification incorporates a new
requirement to verify operator flexibility in determining the method
of verification. Some methods that could be considered as fulfilling
this requirement would include valve alignment checks, or a flow
test verifying a level decrease in the `Q' condensate storage tank
with a corresponding level increase in both steam generators. These
changes result in no significant increase in the consequences of an
accident previously evaluated.
The other proposed changes included in this submittal, including
the Bases changes, are considered to be administrative in nature and
have no effect on the consequences of an accident previously
evaluated. Relocation of the Emergency Feedwater Initiation and
Control (EFIC) requirements from Section 3.4 to Section 3.5 places
the requirements for this instrumentation system with the
requirements for other instrumentation systems, resulting in greater
consistency throughout the ANO-1 TS. Information in the Bases
associated with the EFIC system has been corrected to reflect the
actual plant condition and resolve a conflict with the ANO-1 Safety
Analysis Report. The Bases changes add clarifying information to aid
the operator in determining the applicability of the various EFW
specifications.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed changes introduce no new mode of plant operation.
The EFW system is not an event initiator. It functions to mitigate
the consequences of any event with a loss of normal feedwater.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The changes proposed to the Limiting Conditions for Operation
associated with the EFW system are more conservative than the
current specification, thus resulting in an increase in the margin
of safety. The proposed changes to the actions required when both of
the EFW trains are inoperable and the auxiliary feedwater pump is
unavailable no longer require an immediate plant runback, that is
currently required, which could introduce a plant transient, thus
resulting in an increase in the margin of safety.
The changes revising the Limiting Conditions for Operations also
incorporate and Allowable Outage Time for the turbine-driven EFW
pump steam supply valves which was not previously specified. The 7
day AOT is reasonable based on:
1. The redundant steam supply (from the opposite steam
generator) to the turbine-driven EFW pump is operable,
2. The motor-driven EFW pump is operable, and
3. The probability of an event occurring that would require the
inoperable steam supply valve to actuate is relatively low.
The changes to the surveillance specifications clarify the
proper conditions required for the operability test of the turbine-
driven EFW pump, and revise the requirement for the verification of
proper EFW flow path valve alignment. The change clarifying the test
conditions is required to ensure a sufficient steam supply to the
turbine-driven EFW pump to perform the test. During plant startup,
from an RCS temperature of 280 deg.F to an RCS temperature of
approximately 525 deg.F (corresponding to a steam generator pressure
of approximately 830 psig) the turbine-driven EFW pump is classified
as available until operability is proven by successful completion of
the surveillance requirement. The proposed changes state that the
EFW pumps and their associated flow paths shall be operable when the
RCS is above the cold shutdown condition with any steam generator
relied upon for heat removal (motor-driven EFW pump) and when RCS
temperature is greater than or equal to 280 deg.F (turbine-driven
EFW pump). This specification requires that the flow paths be
properly aligned to maintain operability and is as restrictive as
the current TS 4.8.1.c. The revised specification incorporates a new
requirement to verify proper alignment prior to relying upon any
steam generator for heat removal. This allows the operator
flexibility in determining the method of verification. Some methods
that could be considered as fulfilling this requirement would
include manual valve alignment checks, or a flow test verifying a
level decrease in the `Q' condensate storage tank with a
corresponding level increase in both steam generators.
This change does involve an incremental reduction in the margin
of safety since the extension of the EFW Allowable Outage Time from
36 hours to 72 hours does result in a slight increase in the Core
Damage Frequency (CDF) as calculated in the ANO-1 Probabilistic Risk
Assessment. The new ANO-1 CDF values, incorporating the proposed AOT
extension, are 4.73E-05 (for the turbine-driven EFW pump) and 4.70E-
05 (for the motor-driven EFW pump). These values do not exceed the
NRC Safety Goal of 1.0E-04 per reactor year, as stated in the
Federal Register 50FR32138. The CDF associated with these changes
(6.16E-07 for the turbine-driven EFW pump and 3.04E-07 for the
motor-driven EFW pump) have been evaluated with respect to criteria
contained in SECY-91-270, dated August 27, 1991, and NUMARC 91-04,
dated January 1992, and fall within the category of events of low
risk significance requiring no compensatory measures. This reduction
is not considered significant in that the increase in CDF has been
evaluated as posing no undue risk to the public health and safety
and is categorized as having low risk significance.
The other proposed changes included in this submittal, including
the Bases changes, are considered to be administrative in nature.
Relocation of the Emergency Feedwater Initiation and Control (EFIC)
requirements from Section 3.4 to Section 3.5 places the requirements
for this instrumentation system with the requirements for other
instrumentation systems, resulting in greater consistency throughout
the ANO-1 TS. Information in the Bases associated with the EFIC
system has been corrected to reflect the actual plant condition and
resolve a conflict with the ANO-1 Safety Analysis Report. The Bases
changes add clarifying information to aid the operator in
determining the applicability of the various EFW specifications.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368,
ArkansasNuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County,
Arkansas
Date of amendment request: June 20, 1994
Description of amendment request: The proposed amendments revise
the administrative and control sections of the technical specifications
(TSs) for Arkansas Nuclear One, Units 1 and 2. The proposed changes
relocate controls associated with the ``Review and Audit'' functions
from the TSs to the Quality Assurance Program and relocate requirements
for the audit of emergency and security plans and implementing
procedures from the TSs to the respective emergency and security plans.
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 proposed changes do not affect reactor operations or
accident analyses, have no radiological consequences, and are
considered to be purely administrative in nature. All requirements
relocated from the TSs have been evaluated with respect to the four
criteria of the NRC Final Policy Statement On Technical
Specifications Improvements'' as presented in SECY-93-067, and found
to meet none of the criteria for inclusion in the TS.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed changes introduce no new mode of plant operation
and do not affect the operability of safety-related equipment. All
requirements relocated or deleted from the TSs have been evaluated
with respect to the four criteria of the NRC ``Final Policy
Statement On Technical Specifications Improvements'' as presented in
SECY-93-067, and found to meet none of the criteria for inclusion in
the TS.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
Existing TS operability and surveillance requirements are not
reduced by the proposed change, thus no margins of safety are
reduced. All requirements relocated or deleted from the TSs have
been evaluated with respect to the four criteria of the NRC ``Final
Policy Statement On Technical Specifications Improvements'' as
presented in SECY-93-067, and found to meet none of the criteria for
inclusion in the TS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 9, 1993 as supplemented by
letter dated July 22, 1994
Description of amendment request: The proposed amendment would
revise Section 3.0 and 4.0 of the Technical Specifications (TSs)
consistent with the provision and intent of Generic Letter (GL) 87-09
dated June 4, 1987.
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:
TS 3.0.4 prevents entry into an operational mode or other
specified condition unless Limiting Conditions for Operations (LCOs)
are met without reliance on Action Requirements. The intent of this
TS is to ensure that a higher mode of operation is not entered when
equipment is inoperable or when parameters exceed their specified
limits.
The proposed change clarifies TS 3.0.4 such that LCOs with
Action Statements that permit continued operation for an unlimited
period of time are exempt from the restrictions of TS 3.0.4. This
provision is modified to require an additional plant safety review
prior to implementing additional exceptions to 3.0.4 other than
those currently stated in the individual specifications. This
proposed change is consistent with existing NRC regulatory
requirements for LCOs.
The proposed change to TS 4.0.3 incorporates a 24-hour delay in
implementing the Action Statements due to a missed surveillance
requirement when the Action Statements provide a restoration time
that is less than 24 hours. As reflected in GL 87-09, this change is
justified in that it is overly conservative to assume that systems
or components are immediately inoperable when a surveillance
requirement has not been performed. The NRC concludes in Generic
Letter 87-09 that a 24-hour time limit balances the risks associated
with an allowance for completing the surveillance within this period
against the risks associated with the potential for a plant upset
and challenge to safety systems when the alternative is a shutdown
to comply with Action Statements before the surveillance can be
completed. The NRC further states that the potential for a plant
upset and challenge to safety systems is increased if surveillances
are performed during actions to initiate a shutdown to comply with
Action Requirements.
TS 4.0.4 has been modified to note that its provisions shall not
prevent passage through or to operational modes as required to
comply with Action Requirements. This change is consistent with the
intent of the existing TS and represents a clarification.
No previously analyzed accident scenario is changed by the
proposed TS changes described above. Initiating conditions and
assumptions remain as previously analyzed.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed change to TS 3.0.4 is administrative in nature.
Entry into an operational mode or other specified condition will be
allowed for those specifications not currently stating an exception
to 3.0.4 when 1) the applicable LCOs Action Requirement permits
continued operation for an unlimited period of time and 2) the PORC
[plant operations review committee] has reviewed and approved the
exception.
The proposed change to TS 4.0.3 will allow continued operation
for an additional 24-hours after discovery of a missed surveillance.
As reflected in GL 87-09, missing a surveillance does not mean that
a component or system is inoperable. In most cases, surveillances
provide positive verification of operability.
The proposed change to TS 4.0.4 will alleviate conflict within
the TS. The change is necessary to allow the plant to proceed
through or to required operational modes to comply with Action
Statements even if applicable Surveillance Requirements may not have
been performed.
These changes do not affect the operation of the plant or the
manner in which it is operated.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change to TS 3.0.4 is administrative in nature and
will have no impact on any margin of safety.
The proposed change to TS 4.0.3 will allow up to 24-hours to
perform a missed surveillance. In some cases this will eliminate the
need for a plant shutdown. As reflected in GL 87-09, the overall
effect is an increase in plant safety by avoiding unnecessary
shutdowns and associated system transients due to missed
surveillances.
The proposed change to TS 4.0.4 will eliminate an internal
conflict within the TS and allow the plant to proceed through or to
required operational modes to comply with Action Statements even if
applicable Surveillance Requirements for that mode may not have been
performed.The NRC staff has previously evaluated these change in
Generic Letter 87-09 and determined that the TS modifications will
result in improved TS.
Therefore, the proposed change will 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: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Florida Power and Light Company, et al., Docket No. 50-335, St.
Lucie Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: July 28, 1994
Description of amendment request: The amendment will revise
Technical Specifications (TS) 3/4.4.13 to incorporate Low Temperature
Overpressure Protection (LTOP) requirements similar to those
recommended by the NRC staff via Generic Letter 90-06. The proposed
changes are in accordance with the resolution of Generic Issue 94 for
St. Lucie Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if 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. Each standard is discussed as
follows:
(1)Operation of the facility in accordance with the proposed
amendment would not involve a signifiant increase in the probability
or consequences of an accident previously evaluated.
The changes proposed for St. Lucie Unit 1 Technical
Specifications (TS) 3/4.4.13 are similar to those recommended by the
NRC staff via Generic Letter 90-06 for Low Temperature Overpressure
Protection (LTOP) systems. On the basis of technical studies
performed for Generic Issue 94, the staff concluded that LTOP system
unavailability is a contributor to the risk associated with
overpressure transients during the shutdown modes of plant
operation. Revisions to the actions required and the time for
completion of such actions, in the event that one or more Power
Operated Relief Valves (PORV) become inoperable, provide more rigor
than the existing specifications and are designed to increase LTOP
system availability. The administrative restrictions do not change
the results of existing analyses performed to evaluate postulated
accidents but will improve the availability of systems designed to
mitigate pressure transients that could occur within the LTOP range.
Therefore, operation of the facility in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
(2)Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The changes do
not involve the addition of new equipment or the modification of
existing equipment, nor do they alter the design of St. Lucie plant
systems. Therefore, operation of the facility in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3)Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment provides additional administrative
restrictions for the operation of LTOP equipment. The applicability
of Limiting Conditions of Operation (LCO) involving the PORVs will
be extended to include Operational MODE 6 when the head is on the
reactor vessel, and the rigor of required actions and action
compleiton times in the event that one or more PORVs become
inoperable will be increased. Consequently, the risk of low
temperature operations will be reduced and safety during the
shutdown modes of operation will be enhanced. Therefore, operation
of the facility in accordance with the proposed amendment would not
involve a significant reduction in a margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this
proposed license amendment involves no significant hazards
consideration.
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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Victor M. McCree (Acting)
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: July 1, 1994
Description of amendment request: The proposed amendment to the
Technical Specification (TS) would:1. Modify the facility by providing
an auctioneered power supply for the engineered safety feature
actuation system (ESFAS) sensor cabinets;2 Reinstate the 2-out-of-4
sump recirculation system (SRAS) logic;3. Change Table 3.3 of the
(Safety Feature Actuation System Instrumentation) by adding Manual main
steam isolation (MSI) (Trip Buttons); by removing note (f) which
describes the SRAS logic as a modified 2-out-of-4 logic; and by
replacing Action Statement 4 with an Action Statement that allows
operation with a second inoperable channel, provided both channels are
placed in the bypassed condition. 4. Add to the TS new limiting
conditions for operation and new surveillance requirements together
with BASES (TS 3.3.2.2 and 4.3.2.2.1 and 4.3.2.2.2) for the ESFAS
sensor cabinet power supply drawers.
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 proposed changes do not involve an SHC [significant
hazards consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
SRAS Logic Modification
Implementation of the auctioneered power supply for the sensor
cabinets will permit the reinstatement of the original 2-out-of-4
(six possible combinations) logic for SRAS initiation. The current
logic only has four possible combinations. Changing the minimum
number of SRAS channels required to be operable from four to three
does not significantly reduce the available actuation combinations.
Operation with one channel inoperable will still provide a 2-out-of-
3 logic (three possible trip combinations). With the current SRAS
logic, operation with one channel in bypass does not meet the single
failure criterion for proper SRAS operation. Amendment No. 168
prevents that condition.
Allowing continued operation with three operable channels is
consistent with the original Millstone Unit No. 2 Technical
Specifications (prior to Amendment No. 168).
Note (f), which describes the current logic, will no longer
apply after the auctioneering circuit is installed. This note is for
information only and has no associated action or surveillance
requirements. Therefore, removal of note (f) cannot affect either
the probability or consequences of a postulated accident.
In addition to the change in the minimum number of channels
required to be operable, Action statement 4 will be revised to allow
a limited period of two hours when a second channel may be placed in
bypass for performance of surveillance testing. This is acceptable
due to the installation of the auctioneering circuit and restoration
of the full SRAS logic. Prior to the implementation of the short
term modifications and Amendment No. 168, Action Statement 2 also
applied to the SRAS. That Action Statement allows two hours of
operation with two channels out of service. However, Action
Statement 2 requires one of the two channels to be placed in the
tripped position.
Postulating a LOCA [loss-of-coolant accident] and an additional
failure, while in an action statement that specifies a maximum
allowed outage time, is beyond the design basis of Millstone Unit
No. 2. However, with one SRAS channel in bypass and one in the
tripped position, an additional failure (such as the loss of a DC
vital bus) following the onset of a LOCA could result in a false
SRAS signal.
From an overall safety perspective, the potential consequences
from a false SRAS at the onset of a LOCA are more severe than those
from the failure to automatically generate an actuation signal.
Proposed Action Statement 4 would require actuation of the remaining
channel (following a LOCA and a loss of DC bus as a second failure)
to initiate the SRAS. The existing operation procedures instruct the
operator to ensure that the SRAS actuation occurs when the refueling
water storage tank level decreases to a predetermined value. In the
unlikely event that a LOCA occurred while a Action Statement 4 and
no SRAS was generated at the appropriate time due to an additional
failure which prevents one channel from tripping, the SRAS would be
manually initiated by the operator.
The amount of time that Millstone Unit No. 2 would operate under
Action Statement 4 (with two SRAS channels in bypass) is
approximately 6 hours per month. This is based on the requirement to
conduct monthly channel functional tests for the three operable
channels. The probability of a LOCA occurring during these
surveillance, while in Action Statement 4, with a subsequent failure
of the remaining 2-out-of-2 SRAS logic, is very low.
Sensor Cabinet Auctioneering
The proposed new Technical Specification 3.3.2.2, which
establishes the requirements for the ESFAS sensor cabinets power
supply drawers, permits 48 hours to restore an inoperable sensor
cabinet power supply drawer to operable status. A power supply
drawer renders it inoperable, or if either its normal or backup
power is not available.
Existing Technical Specification 3.8.2.1 contains an 8-hour
action statement for restoring the power sources (VA-10, 20, 30, and
40) if they become inoperable. The proposed 48-hour action statement
for the power supply drawers is appropriate since the sensor cabinet
would remain functional if either normal or alternate power was not
available. However, a LOCA and an additional failure while in the
action statement could result in a false SRAS, since two channels
would supplied from a single DC power supply.
Prior to Amendment No. 168, operation with an inoperable power
supply drawer could continue indefinitely, provided the provisions
of Technical Specification 3/4.3.2 were followed. Operation with a
power supply inoperable for an indefinite period of time places all
the signals associated with that sensor cabinet in the tripped
condition. This creates a 1-out-of-4 tripped condition for SRAS. In
this condition, the single failure required to be postulated could
result in a false SRAS actuation.
This 48-hour action statement is consistent with other action
statements for ESFAS such as Action Statement 1 of Table 3.3-3.
Also, this is consistent with the current wording of Action
Statement 4 which allows 48 hours to restore an inoperable channel
to operable while operating with the modified 2-out-of-4 logic.
MSI Trip Button Addition
The manual trip buttons provide a mechanism for the control room
operator to initiate an MSI trip. The proposed Technical
Specification change will require that a plant shutdown be initiated
if either channel is out of service for more than 48 hours, and
establishes a requirement for surveillance testing every refueling
outage. Including the trip buttons in the Technical Specifications
and establishing operation and surveillance requirements ensures
their operability commensurate with their safety significance.
Based on the above, the changes to Technical Specification 3/4.3
do not increase the probability or consequence of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
SRAS Logic Modification
Changing the number of channels required to be operable from
four to three is acceptable since the original 2-out-of-4 logic will
be restored. This change only affects the number and combinations of
actuation channels necessary to initiate a SRAS. There is no change
to the source or types of initiators, nor is there a change to the
automatic response resulting from a SRAS.
Note (f), which described the modified logic, will no longer
apply after the auctioneering circuit is installed. This note is for
information only and has no associated action or surveillance
requirements. Therefore, removal of note (f) cannot create a new or
different kind of accident.
New Action Statement 4 restores the ability to operate for an
indefinite period of time with one channel in bypass and for a
limited period of time while two channels are out of service. The
change from the original action statement to require that both
channels be in bypass will prevent a false SRAS in the unlikely (and
beyond design basis) event of a LOCA with an additional failure of a
DC bus while in an LCO [limiting condition for operation].
Sensor Cabinet Auctioneering
The addition of a Technical Specification for the sensor cabinet
power supply drawers does not create a potential for a new or
different kind of accident. This new specification implements more
restrictive operating requirements for the sensor cabinets. These
are necessary to ensure that the sensor cabinets are energized from
their primary power supply. The new specification does not affect
the initiation of a SRAS signal nor the type of signal produced.
The auctioneering modification does bring two vital AC
facilities together via isolation devices. This introduces a
potential for a new type of failure mechanism. As described in
Attachment 1, adequate isolation ensures that a failure on one side
of an isolation transformer does not adversely degrade the other
side.
MSI Trip Button Addition
The manual trip buttons provide a mechanism for the control room
operation to initiate an MSI trip. The Technical Specification
change will require that a plant shutdown be initiated if either
manual trip channel is out of service for more than 48 hours, and
establishes a requirement for surveillance testing every refueling
outage. The trip buttons were installed during the 1992 outage.
Establishing operability requirements and surveillance frequency
cannot create a new or different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Involve a significant reduction in the margin of safety.
The net effect of the proposed modifications is to improve the
reliability of the ESFAS and restore the design 2-out-of-4 logic for
the SRAS. The proposed modifications improve the availability of the
ESFAS, and do not affect the vital AC instrument panels.
The Technical Specification changes establish controls for the
used of the SRAS with the restored logic configuration. The
combination of the auctioneering of the power supplies, the
restoration of the 2-out-of-4 logic, and the revised Technical
Specifications restores the margin of safety and operational
flexibility originally designed for the sensor cabinets.
Based on the above, there is no reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
PECO Energy Company, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania
Date of amendment request: June 30, 1994
Description of amendment request: This amendment would remove
certain remote shutdown system control valves and primary containment
isolation valves from Technical Specifications Tables 3.3.7.4-1 and
3.6.3-1 respectively, as a result of eliminating the steam condensing
mode of the Residual Heat Removal 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 Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
These proposed changes will result in abandoning in place
certain remote shutdown system control valves and removing from
service and abandoning in place certain Primary Containment
Isolation Valves (PCIVs) associated with the Residual Heat Removal
(RHR) system steam condensing mode, and will remove the interface
between the High Pressure Coolant Injection (HPCI) and RHR systems,
therefore changing the primary containment pressure boundary.
The RHR system steam condensing mode is a non-safety related
function of the RHR system; however, the pressure and structural
integrity of the associated piping and valves are safety-related.
These proposed changes will not affect any components required to
perform the safety-related function of the RHR or HPCI systems.
The ability of the RHR or HPCI systems to respond to an accident
will not be degraded. Only valves specifically dedicated for use for
the RHR system steam condensing mode will be abandoned in-place, or
removed from the plant. The valves' handswitches which are part of
the remote shutdown panel (RSP) controls, will be physically removed
from the RSP, since they will not perform any function (i.e., the
associated valves will have the electrical power removed). The
flanges and penetration caps that will become part of the primary
containment boundary will be periodically tested for leakage as
required by TS and 10CFR50, Appendix J. All piping and components
that will remain operable will meet the original design
requirements. The other modes of operation of the RHR system (e.g.,
Low Pressure Coolant Injection (LPCI), Shutdown [C]ooling (SDC))
will not be affected by these changes. Therefore, the proposed TS
changes do not involve an increase in 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.
No new failure modes of RHR or HPCI systems are created by the
proposed TS changes. The proposed changes will have no impact on the
existing High Energy Line Break (HELB) analysis for Limerick
Generating Station (LGS). All valves or piping removed and/or
abandoned in place, are dedicated specifically for the RHR system
steam condensing mode, and will not affect the operation of any
components or piping required for other modes of operation of the
RHR or HPCI systems. Therefore, the proposed TS changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The steam condensing mode is a non-safety related function of
the RHR system and, therefore, is not addressed in the TS. This mode
will be physically separated from the other modes of operation of
RHR and HPCI systems, and consequently, will not preclude them from
performing their safety-related functions. The remote shutdown
system control valves to be abandoned in place are not being used
presently, and the proposed changes will not impact the safety
operation of LGS Unit 2. The primary containment penetration caps,
safety-related pipe caps and the flanges replacing the removed PCIVs
will be designed, fabricated and installed in accordance with the
original design requirements, i.e., American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section
III, 1971 Edition with Addenda through Winter 1971. The added
penetration caps and flanges will be capable of maintaining the
primary containment pressure boundary and isolation capabilities
that were required of the PCIVs and will be tested for leakage
periodically, as required by TS and 10 CFR 50, Appendix J.
Additionally, all piping and components that will remain operable
will meet original design requirements. Therefore, the proposed TS
changes do not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Project Director: Charles L. Miller
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment requests: July 19, 1994
Description of amendment request: This amendment will change the
Technical Specification 3.1.5 for each unit for the standby liquid
control system (SLCS) to remove the operability requirement for the
SLCS while in Operational Condition 5 (refueling) with any control rod
withdrawn, and to delete the 18-month system surveillance requirement
(Surveillance Requirement 4.1.5.d.3).
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. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed Technical Specification change to delete the
operability requirement for the SLC System in OPCON 5* (OPERATIONAL
CONDITION 5 with any control rod withdrawn) does not affect the
probability or consequences of an accident previously evaluated.
Design basis accident mitigation scenarios for SSES in OPCON 5 do
not depend on, or require, SLC operability; therefore, the proposed
change to delete SLC operability in OPCON 5* does not affect the
probability or consequences of an accident previously evaluated.
The proposed Technical Specification change to delete
Surveillance Requirement 4.1.5.d.3, 18 month SLC heater operability
check, does not affect the probability or consequences of an
accident previously evaluated. Regarding the SLC heater function,
the operability of the SLC system depends on maintaining the
temperature of the sodium pentaborate solution above 70 deg.F to
prevent the boric acid from precipitating out of solution. SLC
heater 'A' is used to maintain tank temperature between 85 deg. F
and 95 deg.F, thus ensuring that the boric acid remains in solution.
The operability of the heater 'A' is verified through the daily
performance of Technical Specification Surveillance Requirement
4.1.5.a.1, which checks SLC solution temperature, and a control room
alarm. Heater 'B' functions to raise SLC solution temperature prior
to the mixing of SLC chemicals - the mixing of sodium pentaborate
and water is an endothermic (heat consuming) reaction. The
operability of heater 'B' is verified at the time when chemicals are
added to the SLC tank, since a precondition for adding the chemicals
is using heater 'B' to increase tank temperature to 100 deg.F.
Heater 'B' does not function to maintain tank temperature during
normal operation. Therefore, the proposed change does not impact
Susquehanna's ability to maintain SLC solution temperature and thus
does not increase the probability or consequences of an accident
previously evaluated.
2. This proposal does not create the possibility of a new or
different kind of accident or [sic] from any accident previously
evaluated.
The proposed Technical Specification change to delete the
operability requirement for the SLC System in OPCON 5* does not
create the possibility of a new or different kind of accident or
[sic] from any accident previously evaluated. The purpose of the SLC
System is to provide backup capability for bringing the reactor from
full power to a cold, Xenon-free shutdown, assuming that none of the
withdrawn control rods can be inserted. This bases is consistent
with the required operability of the SLC System in OPCONs 1 & 2. The
proposed change does not affect the ability of SLC to meet its
design basis. No credit is taken for SLC in OPCON 5 to mitigate the
effects of reactivity transients, and the SLC system is not designed
to terminate an inadvertent criticality event during core
alterations (OPCON 5) with vessel water level at least 22 feet above
top of vessel flange. Therefore, no new or different accident
scenarios are created by the proposed change.
The proposed Technical Specification change to delete
Surveillance Requirement 4.1.5.d.3, 18 month SLC heater operability
check, does not create the possibility of a new or different kind of
accident or [sic] from any accident previously evaluated. The
proposed change does not affect systems, structures, or components
(SSCs) or the operation of these [SSCs]. The heating and heater
control subsystems of the SLC system will continue to function as
they were designed. The proposed change does not alter the heating
limits or the method for maintaining SLC solution temperature.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident or [sic] from any accident
previously evaluated.3. This change does not involve a significant
reduction in a margin of safety.
The proposed Technical Specification change to delete the
operability requirement for the SLC System in OPCON 5* does not
involve a significant reduction in a margin of safety. The potential
for a decrease in the margin of safety, under this proposed change,
would be associated with periods during OPCON 5* when the SLC system
was not operable. Allowing the SLC system to be inoperable during
OPCON 5* with the vessel level at least 22 feet above top of vessel
flange, represents no reduction in the margin of safety since the
SLC System is not designed to terminate an inadvertent criticality
event with a greater volume of water in the reactor. Having the SLC
system inoperable in OPCON 5* with reactor water levels at normal
operating volumes, does not significantly reduce the margin of
safety because of the number of other design and operating features
which act to prevent inadvertent criticality events. Adequate
shutdown margin is maintained through design and administrative
controls; including, Shutdown Margin Demonstration, Technical
Specification 3.1.1, defueling and refueling procedures, and
refueling interlocks. In addition, the Reactor Protection System
monitors for recriticality and actuates the Control Rod Scram
function if a significant reactivity addition is sensed.
The proposed Technical Specification change to delete
Surveillance Requirement 4.1.5.d.3, 18 month SLC heater operability
check, does not involve a significant reduction in a margin of
safety. Adequate controls are in place, independent of the 18 month
heater operability check, to ensure that the temperature of the
sodium pentaborate solution is maintained above 70 deg. F. These
controls include Surveillance Requirement 4.1.5.a.1, which checks
SLC solution temperature daily, a control room alarm on low and high
temperature, and the ambient temperature conditions in the SLC area
which prevent rapid changes in SLC solution temperature. Operability
of the 'B' heater is not needed to maintain SLC solution
temperature, and the operability of this heater is verified at the
time when chemicals are added to the SLC tank.
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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: Charles L. MillerPower Authority of the State
of New York, Docket No. 50-333, James A. FitzPatrick Nuclear Power
Plant, Oswego County, New York
Date of amendment request: June 13, 1994
Description of amendment request: The proposed amendment would
modify the Facility Operating License by removing License Condition
2.E. This condition applies to the construction cleanup, restoration,
and maintenance of transmission lines. It incorporated into the
Facility Operating License the requirements of U.S. Department of
Interior publication ``Environmental Criteria for Electric Transmission
Systems'' - 1970. The proposed amendment was requested to eliminate
duplication of regulatory authority by government agencies of the same
activity.
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:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will remove a license condition unrelated to
nuclear safety. License condition 2.E incorporated into the
Operating License the requirements of U.S. Department of Interior
publication ``Environmental Criteria for Electric Transmission
Systems'' - 1970. The goal of this standard is to ``safeguard
aesthetic and environmental values within the constraints imposed by
the current state of high-voltage transmission technology.'' License
condition 2.E addresses the preservation of the environment and
natural resources. Removing this condition from the Facility
Operating License has no bearing on plant safety or the health and
safety of the public considering its non-nuclear safety nature. The
transmission line right-of-ways maintained by the Authority are
subject to regulation by other State and Federal agencies. Removal
of this license condition will not affect operation of safety
related structures, systems or components nor affect the quality
assurance program at the FitzPatrick plant. Therefore, the proposed
change does 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.
License condition 2.E of the James A. FitzPatrick Plant
Operating License applies to the construction cleanup, restoration,
and maintenance of transmission lines. The Authority's transmission
lines are managed under guidelines based on the ``Generic
Transmission Line Right-of-Way Management'' plan requirements. The
requirements imposed by the plan on the FitzPatrick transmission
line right-of-ways exceed those of the U.S. Department of Interior
publication referenced in license condition 2.E in both scope and
details. Therefore, implementing the proposed change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. involve a significant reduction in a margin of safety.
License condition 2.E of the James A. FitzPatrick Operating
License applies to the construction cleanup, restoration, and
maintenance of transmission lines. The requirements imposed by this
license condition are unrelated to nuclear safety.
Continued operation of the plant without Condition 2.E does not
involve a significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Michael L. Boyle
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: July 21, 1994
Description of amendment request: The proposed changes would modify
paragraph 2.C.(3) of the Facility Operating License and relocate fire
protection requirements from the Technical Specifications to an
administrative procedure. These changes are based on the guidance
contained in NRC Generic Letter 86-10, ``Implementation of Fire
Protection Requirements,'' and Generic Letter 88-12, ``Removal of Fire
Protection Requirements from Technical Specifications.''
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:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment will not involve a significant hazards
consideration as defined in 10 CFR 50.92, because:
(1) This change does not involve a significant increase in the
probability or consequences of an accident previously evaluated
because no modifications, no changes to operating procedure
requirements, no reduction in administrative controls and no
reduction in equipment reliability are being made as a result of
these changes. This proposed amendment relocates the fire protection
LCOs [Limiting Conditions for Operation] and Surveillance
Requirements from the Technical Specifications to an Administrative
Procedure. No significant changes in content are being made to the
Technical Specification requirements that are being relocated.
Operating limitations will continue to be in effect, and required
surveillances will continue to be performed in accordance with
written procedures and instructions auditable by the NRC.
Although future proposed changes to the fire protection program
elements previously located in the Technical Specifications will no
longer be controlled by 10 CFR 50.36, proposed changes to the Fire
Protection requirements will be controlled by the License Condition
and plant procedures. Programmatic controls will continue to assure
that fire protection program changes do not reduce the effectiveness
of the program to achieve and maintain safe shutdown in the event of
a fire.
(2) The possibility of an accident or malfunction of a different
type than evaluated previously in the safety analysis report is not
created because no reduction to the fire protection requirements, no
modifications, no changes to operating procedure requirements, no
reduction in administrative controls and no reduction in equipment
reliability are being made as a result of these changes.
Programmatic controls will continue to assure that fire protection
program changes do not reduce the effectiveness of the program to
achieve and maintain safe shutdown in the event of a fire.
(3) This proposed amendment does not involve a reduction to the
approved fire protection program or Fire Protection Technical
Specification requirements because the Technical Specification fire
protection requirements are being relocated, with no significant
change in content, to an administrative procedure. Since there is no
reduction in the requirements, no modifications, no changes to
operating procedure requirements, no reduction in administrative
controls and no reduction in equipment reliability are being made as
a result of these changes, there is no 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 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Pao Tsin Kuo
Power Authority of The State of New York, Docket No. 50-286, Indian
PointNuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: July 25, 1994
Description of amendment request: The licensee has requested an
amendment to the Technical Specifications (TS) to revise Table 3.6-1
(Non-Automatic Containment Isolation Valves Open Continuously or
Intermittently for Plant Operation) and Table 4.4-1 (Containment
Isolation Valves) to delete valves SI-1833A(B) and add valves SI-MOV-
1835A(B). The valves being deleted no longer perform a containment
isolation function as a result of a modification which removed the
boron injection tank. The valves being added are needed for testing the
safety injection pumps.
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:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no significant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of any accident
previously evaluated?
Response:
The proposed license amendment does not involve a significant
increase in the probability or consequences of any accident
previously evaluated. The change permits the removal of the two
containment isolation valves on the Boron Injection Tank (BIT)
bypass line. A previous amendment [Amendment No. 139, issued on
October 15, 1993] to the Operating License removed the functional
requirement for the BIT. Consequently, the function of the BIT
bypass line to provide a Safety Injection [SI] pump test flow path
has been rendered obsolete, permitting removal of the bypass line
and associated valves. The bypass line will be cut and capped to
assure containment integrity, therefore eliminating the need for
containment isolation valves SI-1833A and SI-1833B. Opening the BIT
outlet valve [SI-MOV-1835A or B] permits operability testing of the
SI pumps, and is consistent with the current provision permitting
opening of the BIT bypass valves. The changes do not impact the
current operability and surveillance requirements for the Safety
Injection System.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any previously
evaluated?
Response:
The proposed license amendment does not create the possibility
of a new or different kind of accident from any previously
evaluated. The change proposes to eliminate two containment
isolation valves on the BIT bypass line whose function has been
rendered obsolete by a previous amendment to the Operating License.
The bypass line will be cut and capped to assure containment
integrity, therefore eliminating the need for these containment
isolation valves. Intermittent opening of the BIT outlet valve is
consistent with the current provision permitting opening of the BIT
bypass valves, thereby allowing operability testing of the SI pumps.
The changes do not impact the operability or surveillance
requirements for the Safety Injection System.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed license amendment does not involve a significant
reduction in a margin of safety for the following reasons.
Currently, an orientation deficiency with the inboard BIT bypass
isolation valve exposes its stem packing to the non-isolable side of
the valve. The modification corrects this problem by removing both
isolation valves and capping the pipes to assure integrity of the
Containment and Safety Injection System. Additionally, removal of
the isolation valves removes the potential for containment leakage
resulting from valve degradation. Finally, removal of the BIT bypass
line and its associated isolation valves does not inhibit the
ability to test the SI pumps since a previous modification approved
in an Amendment to the Operating License removed the functional
requirement for the BIT. Consequently, the SI pumps may be flow
tested with the BIT inservice, rendering obsolete the function of
the BIT bypass line. Intermittent opening the BIT outlet valve is
consistent with the current provision permitting opening of the BIT
bypass valves, thereby allowing operability testing of the SI pumps.
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 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Pao Tsin Kuo
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: June 29, 1994
Description of amendment request: These proposed amendments would
revise the Technical Specifications to increase the minimum volume of
oil contained in the Diesel Fuel Oil Storage Tanks (DFOSTs) at the
Salem Generating Station (SGS). It would also revise the Updated Final
Safety Analysis Report (UFSAR) description of the fuel oil storage
system capability.
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) [This proposal does] not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Emergency Diesel Generator (EDG) fuel oil is used to support
mitigation of design basis events involving loss of the preferred
(offsite) source of A.C. power. Fuel oil storage capacity has no
effect on the probability of any accident previously evaluated.
Onsite fuel oil storage capability is designed to provide
assurance of long term diesel operation to mitigate the consequences
of a design basis accident. The proposed change would increase the
minimum required volume in the Seismic Class I Diesel Fuel Oil
Storage Tanks (DFOSTs), and would revise the Updated Final Safety
Analysis Report (UFSAR), as part of an effort to reconstitute the
basis for SGS fuel oil storage capacity. The DFOST inventory at the
proposed minimum Technical Specification (TS) limit, combined with
the emergency fill connection and Seismic Class III Fuel Oil Storage
Tank and transfer capability, would continue to provide a long term
onsite fuel oil supply to the EDGs. Operations and Emergency
Preparedness procedures would facilitate the transfer of fuel oil,
and procurement from offsite sources as a contingency measure.
Therefore, the ability to provide a long term supply of fuel oil to
the EDG's is maintained, and the proposed change would not result in
any significant increase in consequences of an accident previously
evaluated.
(2) [This proposal does] not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change would increase the minimum DFOST level
required by TS, and redefines the fuel oil storage and transfer
systems' capability based on plant specific fuel oil consumption
rate and EDG load profiles. These changes would not result in
operation in any configuration prohibited by the present TS, and do
not introduce the possibility of any new type of accident.
(3) [This change does] not involve a significant reduction in a
margin of safety.
The EDG fuel oil storage and transfer capability would continue
to support reliable, long term EDG operation, thereby maintaining an
acceptable margin of safety relative to the ability of onsite A.C.
power to support operation of equipment important to safety. The
proposed changes do not involve a significant reduction in 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: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama
Date of amendment request: March 31, 1994 (TS 319)
Description of amendment request: The proposed amendment revises
the setpoints for instrumentation used to isolate high energy line
breaks in the high pressure coolant injection (HPCI) and reactor core
isolation cooling (RCIC) systems. The proposed amendment also defines
specific areas where steam line space temperatures are monitored.
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes to the HPCI and RCIC steam line space
isolation setpoints do not affect any precursor for any design basis
events or operational transients analyzed in the Browns Ferry Final
Safety Analysis Report. Therefore, the probability of an accident
previously evaluated is not increased.
The HPCI and RCIC steam line space high temperature isolations
are provided to ensure automatic closure of each system's primary
containment isolation valves for a HPCI or RCIC steam line break.
The isolation occurs when a very small leak has occurred. If the
small leak is allowed to continue without isolation, offsite dose
limits may be reached. As a result of the environmental
qualification program, the environmental responses of the reactor
building to high energy line breaks were analyzed. TVA used computer
modeling techniques to predict the temperature response of various
reactor building zones to high energy line breaks. The results
indicate that the setpoints for the HPCI and RCIC temperatures
switches should be lowered. The lower setpoints assure the timely
initiation of a closure signal to the primary containment isolation
valves. Therefore, assuring the maximum allowable temperatures are
not exceeded.
The proposed change to the HPCI and RCIC steam line space
isolation setpoints are in the conservative direction and provides
the same or earlier detection and isolation of HPCI and RCIC steam
line breaks.
The proposed trip level settings are high enough to ensure that
spurious trips do not occur from normal or transient system
operation and low enough to ensure that line breaks are detected and
isolated before design conditions are exceeded. Therefore, the
proposed changes will not significantly increase the consequences of
an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change to the HPCI/RCIC steam line space high
temperature isolations does not involve any modification to plant
equipment or changes in operating procedures. No new failure modes
are introduced. There is no effect on the function or operation of
any other plant system. No new system interactions have been
introduced by the change. The results of a break in the HPCI or RCIC
steam lines remain as before. The HPCI or RCIC steam line area
temperature switches will still detect a break due to an increase in
area temperature and provide an initiation signal to close the
system primary containment isolation valves to prevent reactor
coolant loss. The proposed change will conservatively serve to
detect and mitigate HPCI and RCIC line breaks more expeditiously.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not reduce the margin of safety. The
proposed change ensure that HPCI and RCIC steam line breaks are
isolated at the same or lower steam line area temperatures. Computer
modeling techniques were utilized to predict the temperature
response in various areas through which the HPCI and RCIC steam
lines pass. The revised setpoints are established above the maximum
expected normal room temperatures to avoid spurious actions due to
ambient conditions and below the analytical limits to ensure timely
pipe break detection and isolation. Substantial margin exist between
the maximum temperature expected in each area and the minimum
actuation temperature determined for each temperature switch. With
the substantial margin between maximum temperatures for the areas
and the minimum actuation temperature of the switches, the maximum
temperatures cannot result in actuation of the switches. The design
and function of the affected components has not been changed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Mr. Frederick J. Hebdon
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of amendment request: May 11, 1994 (TS 347T)
Description of amendment request: The proposed amendment extends
the allowed outage time for the Browns Ferry Nuclear Plant (BFN) Unit 2
250 volt DC (direct current) control power supplies from 5 to 45 days.
The amendment is a temporary revision to the BFN Unit 2 Technical
Specifications (TS) to permit replacement of batteries and other
hardware.
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change involves temporarily (one-year period)
extending the 5-day AOT [allowed outage time] for the 250-volt
shutdown board control power supplies to 45 days. As such, this
change does not increase the probability of any accident previously
analyzed.
The 250-volt DC Power System is required to function to mitigate
the consequences of design basis accidents. The loss of a single
250-volt DC shutdown board control power supply will result in a
loss of control power for the 480-volt and the 4160-volt shutdown
board that it serves. Loss of control power results in loss of only
those engineered safeguards supplied by its respective shutdown
boards. Redundant safe shutdown equipment exists to mitigate the
consequences of design basis accidents. As discussed in Final Safety
Analysis Report (FSAR) subsection 8.6.4.3, a single failure of a
shutdown board control power supply is acceptable.
Loss of a single 250-volt plant DC power supply will not prevent
Unit 2 safe shutdown. The 250-volt plant DC power supply system is
designed so that any two out of the three power supplies carry the
entire load needed for safe shutdown. As discussed in FSAR
subsection 8.6.4.2 a single failure of a 250-volt plant DC power
supply is acceptable.
At no time will control power be unavailable to the shutdown
boards during the system upgrades. The proposed change will only
increase the time allowed to operate the plant while a 250-volt DC
shutdown board control power supply is out of service.
The proposed TS change allows an additional 40 days to perform
system upgrades and results in a small increase in risk. This small
increase in risk is associated with the probability and consequences
of a 250-volt plant DC power supply malfunction while it is
supplying shutdown board control power. The increase in risk
associated with extending the AOT was analyzed in a Probabilistic
Safety Assessment (PSA) and determined to be approximately 0.3
percent. This small increase in risk is determined to be
insignificant and well within the uncertainty bounds of the PSA.
The proposed TS change does not change the function of any plant
structure, system or component. The proposed change allows for
improvements to the 250-volt DC shutdown board control power supply
system. The improvements will increase the capability and
reliability of the system. Qualified backup power will be utilized
at all times during system modifications. Only one power supply will
be out of service at a time during the modifications.
The small increase in risk is more than offset by the increased
capability, capacity, and reliability of the new power supplies.
Therefore, the power supply modifications will result in a net
overall safety benefit.
[The licensee has also committed to implement compensatory
measures while performing the power supply modifications. These
measures provide additional confidence that potential accident
consequences are not increased.]
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Extending the 5-day AOT for the 250-volt shutdown board control
power supplies to 45 days does not create the possibility of a new
or different kind of accident, nor does it increase the probability
that an accident will occur. The AOT extension does not involve
plant modifications that could create the possibility of a new or
different kind of accident from any of those discussed in the FSAR.
The 250-volt DC shutdown board control power supply
modifications involve replacement of the existing components with
more reliable, increased capacity equipment having the same
functions as before.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed TS change involves a risk increase of approximately
0.3 percent. TVA [the Tennessee Valley Authority, the licensee]
considers this small increase to be insignificant. TVA also
considers that the small increase in risk is offset by the benefits
associated with replacing the control power supplies with new,
upgraded equipment. Therefore, the proposed TS change does not
involve a significant reduction in a margin of safety.
[The licensee has also committed to implement compensatory
measures while performing the power supply modifications. These
measures provide additional capability to mitigate an accident,
minimizing any effect on safety margin.]
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Mr. Frederick J. Hebdon
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: July 18, 1994
Description of amendment request: The proposed amendment would
modify Point Beach Nuclear Plant Technical Specification (TS) 15.3.7,
``Auxiliary Electrical Systems,'' by including an allowed outage time
for one of the four connected station battery chargers and subsequent
shutdown requirements. The basis for Section 15.3.7 would also be
revised to support the above changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
In accordance with the requirements of 10 CFR 50.91(a),
Wisconsin Electric Power Company (Licensee) has evaluated the
proposed changes against the standards of 10 CFR 50.92 and has
determined that the operation of Point Beach Nuclear Plant, Units 1
and 2, in accordance with the proposed amendments, does not present
a significant hazards consideration. A proposed facility operating
license amendment does not present a significant hazards
consideration if operation of the facility in accordance with the
proposed amendment will not:
1. Create 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. Will not create a significant reduction in a margin of
safety.
The proposed amendment allows operation for up to two hours with
one out of the four connected station battery chargers out of
service. The 2-hour outage time is based on Regulatory Guide 1.93
and reflects a reasonable time to assess plant status and either
connect an operable battery charger to the affected DC bus or
prepare to effect an orderly and safe shutdown of the operating
unit(s). Since the batteries, chargers, and their associated vital
instrument buses provide sufficient redundancy to assure the
initiation of proper protective actions during degraded system
conditions, operation of PBNP in accordance with these proposed
amendments cannot create an increase in the probability or
consequences of an accident previously evaluated, create a new or
different kind of accident, or result in a significant reduction in
a margin of safety. Therefore, the proposed changes do not present a
significant hazards consideration.
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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon
Previously Published Notices Of consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Georgia Power Company, Docket No. 50-366, Edwin I. Hatch Nuclear
Plant, Unit 2, Appling County, Georgia
Date of amendment request: July 19, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 3.3.6.6 to permit the traversing incore
probe (TIP) system to be considered operable with less than four
operable TIP units.Date of publication of individual notice in Federal
Register: July 22, 1994 (59 FR 37516) Expiration date of individual
notice: Comment Period expires August 8, 1994; Notice period expires
August 22, 1994
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
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 20555, and at the local public document
rooms for the particular facilities involved.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: September 15, 1993
Brief description of amendment: The amendment revises the pressure-
temperature limits from 15 to 24 effective full power years.
Date of issuance: July 29, 1994
Effective date: July 29, 1994
Amendment No. 149
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 13, 1993 (58 FR
52980) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 29, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
Home and Fifth Avenues, Hartsville, South Carolina 29550
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: February 4, 1994
Brief description of amendment: The amendment revises the Action
Statement of TS 3.6.5, Vacuum Relief System, to require in Modes 1-4
with one vacuum relief system inoperable that the system be restored to
the operable status within seventy-two hours or be in at least hot
standby within the next six hours.
Date of issuance: July 27, 1994
Effective date: July 27, 1994
Amendment No. 49
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14886) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 27, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: May 11, 1994
Brief description of amendment: The amendment revises TS 3/4.2.3 to
establish limits on reactor power level as a function of total reactor
coolant system (RCS) flow rate up to 5 percent below the current
specified flow rate.
Date of issuance: July 27, 1994
Effective date: July 27, 1994
Amendment No. 50
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27079) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 27, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of application for amendments: March 30, 1994, as supplemented
by letters dated June 13, June 14, July 11, July 21 and July 28, 1994
Brief description of amendments: The amendment revises the
Technical Specifications (TSs) by changing the Unit 1 heatup and
cooldown pressure-temperature (P-T) curves (i.e., Figures 3.4-2a and
3.4-3a) to incorporate a newly determined reactor pressure vessel (RPV)
reference nil-ductility temperature, RTNDT. This new value of
RTNDT was determined from the licensee's analysis of the first
irradiation sample removed from Unit 1. The setpoint curve contained in
Figure 3.4-4a for the Unit 1 Low Temperature Overpressure Protection
System (LTOPS) is also revised to reflect the changes in the P-T curves
and to provide a margin for uncertainties in measuring the reactor
pressure. Additionally, the amendment updates the removal schedule of
RPV surveillance capsules for both units in accordance with the
American Society for Testing Materials (ASTM) Standard ASTM E185-82.
Finally, the amendment incorporates an editorial change for Unit 2 in
which some clarifying text was added in the Table of Contents to
indicate the lifetime applicability of Figure 3.4-4b for Unit 2.
Date of issuance: July 29, 1994
Effective date: July 29, 1994
Amendment Nos.: 53 and 53
Facility Operating License Nos. NPF-72 and NPF-77. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 12, 1994 (59 FR
24747) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 29, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: September 10, 1993 as
supplemented November 17, 1993
Brief description of amendments: The amendments revise the LaSalle
County Station, Units 1 and 2 Updated Final Safety Analysis Report
Section 11.5.2.1.4 to specify that operator action is required to trip
the mechanical vacuum pump upon receipt of a main steam line high
radiation alarm, rather than the action of an automatic trip, which is
currently described in the UFSAR. NRC approval was required because the
required operator action, an existing condition, is contrary to that
described in the UFSAR and the NRC's Safety Evaluation Report related
to the operation of LaSalle County station (NUREG-0519), and involved
an unreviewed safety question.
Date of issuance: July 26, 1994
Effective date: July 26, 1994
Amendment Nos.: 101 and 85
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the UFSAR.
Date of initial notice in Federal Register: December 1, 1993 (58 FR
63403) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 26, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of application for amendments: June 16, 1994
Brief description of amendments: The amendments change
specification 3/4.10.1 to recognize the exemption of a single valve on
each unit from Type C testing until the next refueling outage on each
unit.
Date of issuance: August 1, 1994
Effective date: August 1, 1994
Amendment Nos.: 155 and 143
Facility Operating License Nos. DPR-39 and DPR-48. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 30, 1994 (59 FR
33798) Public comments requested as to proposed no significant hazards
consideration: yes. The notice provided an opportunity to submit
comments on the Commission's proposed no significant hazards
consideration determination. No comments have been received. The notice
also provided an opportunity to request a hearing by August 1, 1994,
but indicated that if the Commission makes a final no significant
hazards consideration determination, any such hearing would take place
after issuance of the amendment. The Commission's related evaluation of
the amendment and final significant hazards consideration determination
is contained in a Safety Evaluation dated August 1, 1994.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Consolidated Edison Company of New York, Docket No. 50-003 and
Docket No. 50-247, Indian Point Nuclear Generating Unit Nos. 1 and
2, Westchester County, New York
Date of application for amendments: September 29, 1993
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) to change the submittal frequency of the
Radioactive Effluent Release Report from semiannually to annually, and
change the reporting date.
Date of issuance: July 21, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 44 and 172
Facility Operating License Nos. DPR-5 and DPR-26: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62153) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 21, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
PointNuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: September 29, 1993, as
supplemented by letter dated April 1, 1994.
Brief description of amendment: The amendment revises the Technical
Specifications to remove the cycle-specific parameter limits and to
reference a Core Operating Limits Report containing these limits. These
changes are in accordance with the guidance provided in Generic Letter
88-16, ``Removal of Cycle-Specific Parameter Limits from Technical
Specifications.''
Date of issuance: July 26, 1994
Effective date: This license amendment is effective as of the date
of issuance of the COLR by the licensee to be implemented no later than
the return to operation following the 1995 refueling outage.
Amendment No.: 173
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62154) The April 1, 1994, provided additional information that did
not change the initial determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
July 26, 1994.No significant hazards consideration comments received:
No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.Consolidated Edison
Company of New York, Docket No. 50-247, Indian PointNuclear Generating
Unit No. 2, Westchester County, New York
Date of application for amendment: January 28, 1994
Brief description of amendment: The amendment revises the TSs to
change the containment isolation valve testing frequency and the
acceptance criteria for the combined containment leakage rate to
accommodate operation on a 24-month fuel cycle. These changes follow
the guidance provided in Generic Letter 91-04, ``Changes in Technical
Specification Surveillance Intervals to Accommodate a 24-Month Fuel
Cycle.''
Date of issuance: July 29, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 174
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17596) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 29, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of application for amendment: April 22, 1994, as supplemented
July 6, 1994
Brief description of amendment: The amendment revised the reactor
vessel pressure-temperature limits in the Technical Specifications. The
change insures that the vessel fracture toughness requirements of
Section V of 10 CFR Part 50, Appendix G, are satisfied through end of
life.
Date of issuance: July 25, 1994
Effective date: July 25, 1994
Amendment No.: 113
Facility Operating License No. DPR-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1994 (59 FR
24749). The July 6, 1994, letter provided clarifying information within
the scope of the initial notice and did not affect the staff's proposed
no significant hazards consideration findings. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
July 25, 1994.No significant hazards consideration comments received:
No.
Local Public Document Room location: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: May 10, 1994
Brief description of amendment: The amendment revises the Fermi-2
Technical Specifications (TS) to remove Table 3.6.3-1, the list of
primary containment isolation valves and Table 3.8.4.3-1, the list of
safety systems' motor-operated valve thermal overload protection from
the TS to administrative procedures in accordance with the guidance
contained in Generic Letter 91-08.
Date of issuance: August 1, 1994
Effective date: August 1, 1994, with full implementation within 45
days.
Amendment No.: 102
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29626) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 1, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: November 11, 1993, as
supplemented on June 13, 1994
Brief description of amendments: The amendments revise the
Technical Specification surveillance requirements for the emergency
core cooling system subsystems.
Date of issuance: July 29, 1994
Effective date: July 29, 1994
Amendment Nos.: 145 and 127
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17579) The June 13, 1994, letter provided clarifying and additional
information that did not change the scope of the November 11, 1993,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 29, 1994. No significant hazards
consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: May 5, 1994, as supplemented
June 13, 1994.
Brief description of amendments: The amendments revise the
Technical Specifications to increase Main Steam and Pressurizer Code
Safety Valve Setpoint Tolerances.
Date of issuance: August 2, 1994
Effective date: August 2, 1994
Amendment Nos.: 146 and 128
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 1994 (59 FR
32029) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 2, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 14, 1993
Brief description of amendment: The amendment revised the Technical
Specifications to revise the azimuthal power tilt limit from less than
or equal to 0.10 (10%) to less than or equal to 0.03 (3%) and revises
the action statement for control element assembly misalignment to allow
24 hours to restore the tilt to less than 3%.
Date of issuance: August 3, 1994
Effective date: August 3, 1994
Amendment No.: 97
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2866) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 3, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: April 6, 1994, as supplemented
June 21, 1994
Brief description of amendment: The amendment eliminates the scram
and main steam line isolation valve (MSIV) closure requirements
associated with the main steam line radiation monitors (MSLRM). The
amendment also eliminates the following related automatic isolation
functions that are associated with the MSLRM scram and MSIV isolation:
a) Main Steam Line Condenser Drain Valves, b) Emergency Condenser Drain
Valves, c) Reactor Recirculation Loop Sample Valve, d) Instrumental Air
Valves, and e) Condenser Pump Isolation.
Date of issuance: July 29, 1994
Effective date: As of the date of issuance to be implemented at the
restart from refueling outage 15R.
Amendment No.: 169
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22008). The June 21, 1994, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of
this amendment is contained in a Safety Evaluation dated July 29,
1994.No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: July 15, 1993.
Brief description of amendment: The amendment revises the plant
Technical Specifications (TSs) on the Reactor Coolant Inventory
Trending System (RCITS). The change is consistent with NUREG-1430
entitled ``Standard Technical Specifications for Babcock and Wilcox
Plants.'' The RCITS information will be available to the operator to
enhance the operator's ability to understand and manage transients and
events when needed.
Date of issuance: August 1, 1994
Effective date: As of its date of issuance, to be implemented
within 30 days of issuance.
Amendment No.: 191
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications. I11Date of initial notice in Federal
Register: June 8, 1994 (59 FR 29626) The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
August 1, 1994. No significant hazards consideration comments received:
No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: February 10, 1994
Brief description of amendment: The amendment revises the TMI-1
Technical Specifications (TS) to revise specification 3.7.2.c, ``Unit
Electric Power System,'' to provide an option to testing an emergency
diesel generator (EDG) when the redundant EDG is inoperable.
Date of issuance: July 25, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 188
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32230) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated July 25, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: March 2, 1994
Brief description of amendment: The amendment revises the plant
Technical Specifications to modify Operational Safety Instrumentation
requirements to specify completion time which allows for performance of
maintenance or surveillance within a reasonable time and to be
consistent with the allowable outage time for other safety-related
equipment when only one train is affected.
Date of issuance: July 25, 1994
Effective date: As of the date of issuance to be implemented within
30 days after issuance.
Amendment No.: 189
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17600). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 25, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: March 11, 1994
Brief description of amendment: The amendment revises the plant
Technical Specifications to specify an allowable outage time for the
Emergency Feedwater Pumps during surveillance activities. It also
changes the requirement to test redundant components for operability to
a requirement to ensure operability based on verification of completion
of appropriate surveillance activities.
Date of issuance: July 25, 1994
Effective date: As of the date of issuance to be implemented within
30 days of issuance.
Amendment No.: 190
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17601).The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 25, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: January 14, 1994
Brief description of amendment: The amendment revised TS Sections
3/4.3, ``Instrumentation,'' 3/4.4.2, ``Safety/Relief Valves,'' and
associated Bases to increase the surveillance test intervals and
allowable out-of-service times for various instruments.
Date of issuance: August 2, 1994
Effective date: August 2, 1994
Amendment No.: 74
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 26, 1994 (59 FR
21787) The additional information contained in the supplemental letter
dated July 15, 1994, was clarifying in nature and thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination.The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated August 2, 1994.No significant hazards consideration comments
received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: March 16, 1994
Brief description of amendments: The amendments modified Figure
3.4-4, ``Nominal Maximum Allowable PORV Setpoint for the Cold
Overpressure System,'' for the cold overpressure mitigation system with
a revised setpoint curve.
Date of issuance: August 3, 1994
Effective date: August 3, 1994, to be implemented within 31 days of
issuance
Amendment Nos.: Unit 1 - Amendment No. 63; Unit 2 - Amendment No.
52
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17601) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 3, 1994.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
Northeast Nuclear Energy Company, Docket No. 50-245,
MillstoneNuclear Power Station, Unit 1, New London County,
Connecticut
Date of application for amendment: April 29, 1994
Brief description of amendment: The amendment changes the
requirement for reactor operators in Table 6.2-1 from 2 to 3 for the
RUN, STARTUP/HOT STANDBY and HOT SHUTDOWN conditions. In addition, two
typographical corrections were made to page 6-4.
Date of issuance: August 2, 1994
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 75
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32231) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 2, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay
Power Plant, Unit 3, Humboldt County, California
Date of application for amendment: October 8, 1993 (Reference HBL-
93-058)
Brief description of amendment: This amendment modified the
Technical Specifications (TS) incorporated in
Facility Operating License No. DPR-7 as Appendix A by incorporating
a title change into Section VII, Administrative Controls. This change
reflects a plant organizational name change.
Date of issuance: July 26, 1994
Effective date: This license amendment is effective as of the date
of its issuance and must be fully implemented no later than 30 days
from the date of issuance.
Amendment No.: 27
Facility Operating License No. DPR-7: The amendment revised the TS.
Date of initial notice in Federal Register: January 5, 1994 (59 FR
624) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 26, 1994No significant hazards
consideration comments received: No.
Local Public Document Room location: Humboldt County Library, 636
F Street, Eureka, California 95501
Philadelphia Electric Company, Public Service Electric and Gas
CompanyDelmarva 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: October 27, 1993, as
supplemented by letters dated April 29, 1994, and June 27, 1994
Brief description of amendments: These amendments revise the Unit 2
and Unit 3 Technical Specifications to allow one of the required on-
shift senior reactor operators (SRO) to be combined with the required
shift technical advisor (STA) position (i.e., dual-role SRO/STA
position) as long as a minimum of three qualified individuals fill the
SRO and STA positions.
Date of issuance: August 2, 1994
Effective date: August 2, 1994
Amendments Nos.: 191 and 196
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64613)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 2, 1994.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,
Pennsylvania 17105.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: May 10, 1994
Brief description of amendment: The Technical Specifications
amendment revised Section 3.1.C.3 and Table 4.1-1 of Appendix A of the
Operating License. These changes require that the reactor coolant
average temperature (Tavg) be no lower than 540 deg.F during
critical operation. Critical operation at Tavg less than 540 deg.F
requires operator response to restore Tavg to greater than or
equal to 540 deg.F within 15 minutes or be in hot shutdown within the
following 15 minutes. Additionally, the change in Table 4.1-1 entitled,
``Minimum Frequencies for Checks, Calibrations and Tests,'' adds the
requirement for Tavg instrument check frequency to be reduced to
30 minutes when the Tavg banks are above zero steps. Furthermore,
the revision to the Bases indicates that the minimum temperature for
criticality provides assurance that the reactor is operated within the
bounds of the safety analyses. Also included is an administrative
change to correct some typographical errors on page 3.1-25 of the
Technical Specifications.
Date of issuance: July 25, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 149
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29630) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 25, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: May 3, 1994
Brief description of amendments: The amendments revised the
Technical Specification (TS) for Combustible Gas Control (3/4.6.4.1) by
changing the surveillance frequency for performing the channel
functional test to once-per-quarter and the channel calibration to
once-per-refueling. Also, the TS for the Auxiliary Feedwater System (3/
4.7.1.2) were changed to reduce the surveillance frequency for
performing pump operability tests to once-per-quarter on a staggered
test basis. These changes are consistent with the provisions of Generic
Letter 93-05, ``Line-Item Technical Specifications Improvements to
Reduce Surveillance Requirements For Testing During Power Operations.''
Date of issuance: July 27, 1994
Effective date: July 27, 1994
Amendment Nos. 153 and 134
Facility Operating License Nos. DPR-70 and DPR-75. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29634) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 27, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: March 11, 1994
Brief description of amendment: The amendment changes the Technical
Specifications to delete TS Surveillance Requirement 4.8.4.1.a.3 that
requires periodic retest of containment penetration overcurrent
protection fuses and to remove references to containment penetration
fuse testing from the TS Bases.
Date of issuance: July 29, 1994
Effective date: July 29, 1994
Amendment No.: 115
Facility Operating License No. NPF-12. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1994 (59 FR
24752) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 29, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180.
Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama.
Date of application for amendment: June 17, 1994
Brief description of amendment: The amendment changes the Technical
Specifications to revise the nuclear enthalpy rise hot channel factor
(F delta H) from equal to or less than 1.65 [1 plus 0.3(1-P)] to equal
to or less than l.70 [1 plus 0.3(1-P)] where P is a fraction of rated
power. The amendment also revises the action statement to reflect
guidance contained in the improved standard technical specifications.
Date of issuance: July 22, 1994
Effective date: July 22, 1994
Amendment No.: 109
Facility Operating License No. NPF-2. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32249) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 22, 1994.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 36302
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama
Date of application for amendments: August 25, 1992 (TS 321)
Brief description of amendments: The amendments delete reference to
recirculation equalizer valves from the technical specifications. These
components have been removed from Browns Ferry Unit 3, and are not used
in Browns Ferry Units 1 and 2.
Date of issuance: August 4, 1994
Effective date: August 4, 1994
Amendment Nos.: 211, 226 and 184
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 28, 1992 (57 FR
48829) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 4, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: June 17, 1993 (TS 93-08)
Brief description of amendments: The amendments revise the
allowable values for the intermediate and source range neutron flux
reactor trip setpoints.
Date of issuance: July 26, 1994
Effective date: July 26, 1994
Amendment Nos.: 185 and 177
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41514) The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated July 26, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: February 17, 1993
Brief description of amendment: This amendment increases the TS
trip setpoint and its associated allowable value for containment high-
radiation specified in TS Table 3.3-4 from ``<2 x Background at RATED
THERMAL POWER'' to ``<4 x Background at RATED THERMAL POWER.''
Date of issuance: July 27, 1994
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No. 190
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34096) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 27, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: February 14, 1994, as supplemented by
letter dated April 29, 1994.
Brief description of amendments: The amendments increase the boron
concentration limits for the Unit 2 refueling water storage tank and
emergency core cooling system, and delete a footnote concerning
refueling canal boron concentration during initial fuel load for both
units.
Date of issuance: August 2, 1994
Effective date: August 2, 1994, to be implemented prior to startup
for Cycle 2 for Comanche Peak Steam Electric Station, Unit 2.
Amendment Nos.: Unit 1 - Amendment No. 26; Unit 2 - Amendment No.
12
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22015) The information contained in the April 29, 1994, letter was
editorial in nature and thus, within the scope of the initial notice
and did not affect the staff's proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated August 2, 1994.No
significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: February 17, 1994, supplemented
by letter dated May 18, 1994
Brief description of amendment: The amendment revises the Technical
Specification 3/4.5.1 and associated Bases Section 3/4.5.1. A new
Action Statement a. provides a 72-hour allowed outage time (AOT) for
one accummulator inoperable due to boron concentration. The Action
Statement b. AOT was changed to 24 hours. Surveillance Requirements
4.5.1.1.a.1 and 4.5.1.1.b were revised and 4.5.1.2 was deleted from the
TS.
Date of issuance: August 5, 1994
Effective date: August 5, 1994 to be implemented within 30 days
Amendment No.: 91
Facility Operating License No. NPF-30. Amendment revised the
Technical Specifications 3/4.5.1 and associated Bases Section 3/4.5.1.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14898) The additional information contained in the May 18, 1994, letter
provided additional supplemental information that did not change the
initial proposed no significant hazards consideration. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated August 5, 1994.No significant hazards consideration comments
received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: April 19, 1994
Brief description of amendments: The amendments revise the NA-1&2
Technical Specifications surveillance frequency requirements for
control rod motion testing from once per 31 days to once per 92 days.
Date of issuance: July 28, 1994
Effective date: July 28, 1994
Amendment Nos.: 185 and 166
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27070) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 28, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
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 application for amendments: April 19, 1994
Brief description of amendments: These amendments modify the
surveillance frequency of the control rod motion testing from monthly
to quarterly
Date of issuance: August 2, 1994
Effective date: August 2, 1994
Amendment Nos. 192 and 192
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27070) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 2, 1994No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: April 1, 1993
Brief description of amendment: The amendment modifies the
Technical Specifications to add inservice inspection requirements for
reactor coolant system piping in accordance with Generic Letter 88-01,
``NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in
BWR Austenitic Stainless Steel Piping.'' In addition, the amendment
corrects an administrative error in a TS that references a table
listing high/low pressure interface valve leakage pressure monitors.
Date of issuance: July 28, 1994
Effective date: 30 days after the date of issuance.
Amendment No.: 130
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1993 (58 FR
28065) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 28, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: February 17, 1994, supplemented
by letter dated May 13, 1994.
Brief description of amendment: The amendment changes the Technical
Specifications (TS) 10-year hydrostatic testing requirements. The
changes (1) add a special test exception for inservice leak testing and
hydrostatic testing, (2) add a new minimum reactor vessel metal
pressure-temperature curve for less than or equal to eight effective
full power years, and (3) delete Table B 3/4.4.6-1, ``Reactor Vessel
Toughness,'' from the TS bases.
Date of issuance: May 27, 1994
Effective date: May 27, 1994
Amendment No.: 122
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14902) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 27, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
NuclearPower Plant, Kewaunee County, Wisconsin
Date of application for amendment: December 1, 1993
Brief description of amendment: The amendment revises the Kewaunee
Nuclear Power Plant (KNPP) Technical Specifications (TS) by
incorporating technical and administrative changes to TS 3.10, Control
Rod and Power Distribution Limits. The changes eliminate specifications
for fuel designs no longer used at Kewaunee, specify required actions
to be taken upon exceeding control bank insertion limits, and revise
the limits for Departure from Nucleate Boiling (DNB) related parameters
to assure operation within the assumptions of the Updated Safety
Analysis Report (USAR) analyses.
Date of issuance: August 3, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 110
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4949) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 3, 1994. No significant hazards
consideration comments received: No.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By September 16, 1994, 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 20555 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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendments: August 2, 1994
Brief description of amendments: The amendments revise the
Technical Specifications by adding a footnote that recognizes that
through the end of cycle 6, the Unit 1, loop B wide range hot leg
indication at the remote shutdown panel is inoperable.
Date of issuance: August 5, 1994
Effective date: August 5, 1994
Amendment Nos.: 63 and 63
Facility Operating License Nos. NPF-37 and NPF-66. The amendments
revised the Technical Specifications.Public comments requested as to
proposed no siginificant hazards consideration: No. The Commission's
related evaluation of the amendments, finding of emergency
circumstances, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated August 5,
1994.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
Local Public Document Room location: Byron Public Library, 109 N.
Franklin, P.O. Box 434, Byron, Illinois 61010.
NRC Project Director: Robert A. Capra
Dated at Rockville, Maryland, this 10th day of August 1994.
For The Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor
Regulation
[Doc. 94-20006 Filed 8-16-94 8:45 am]
BILLING CODE 7590-01-F