[Federal Register Volume 60, Number 227 (Monday, November 27, 1995)]
[Notices]
[Pages 58394-58416]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-28606]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from October 28, 1995, through November 9, 1995.
The last biweekly notice was published on Wednesday, November 8, 1995
(60 FR 56361).
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.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By December 27, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any
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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, 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, and at the local public document room for
the particular facility involved.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: October 20, 1995.
Description of amendment request: The proposed one-time amendment
would revise the Calvert Cliffs Nuclear Power Plant, Unit No. 1, (CC-1)
Technical Specifications (TSs) by extending certain 18-month instrument
surveillance intervals by a maximum of 39 days to March 31, 1996. The
instruments involved are included in the reactor protective system,
engineered safety features actuation system, power-operated relief
valves, low-temperature overpressure protection system, remote shutdown
instruments, post-accident monitoring, radiation monitoring, and
containment sump level instruments.
The Commission issued Amendment No. 208 to Facility Operating
License No. DRP-53 and Amendment No. 186 to Facility Operating License
No. DRP-69 for the CC-1/2, respectively. The amendments permanently
extended the surveillance intervals for the instruments described above
from 18 months to 24 months after a specified number of the instruments
had been replaced. The amendments were effective immediately and to be
implemented on CC-2 within 30 days, but not implemented on CC-1 until
its restart after the spring 1996 refueling outage. All of the
instruments identified for replacement on CC-2 have been replaced, but
those identified for replacement on CC-1 have not been replaced, thus,
the reason for the later implementation date. The proposed one-time
amendment is needed prior to Amendment No. 208 being implemented
because of a change in the refueling schedule. The licensee has
provided technical justification to allow operation for an additional
short-time period of up to a maximum of 39 days.
CC-1 was initially scheduled to begin its refueling outage on
February 16, 1996, which would have been within the time frame
necessary to perform the required 18-month instrument surveillances
currently required for the instruments identified above. The licensee
has recently rescheduled the refueling outage for CC-1 to start March
15, 1996, several months after the initial amendment request and after
consultation with the Pennsylvania-New Jersey-Maryland power pool. The
revised schedule will allow the maximum use of the available fuel in
the CC-1 reactor core and will also allow the unit to operate for an
additional period of about 1 month during a period of potentially high
power demand. In addition, the delay will allow more time to plan and
prepare for the upcoming refueling outage. Performing the required
instrument surveillances at power would present an unwarranted
personnel safety risk and, in some cases, the surveillances cannot be
done during power operation because they would cause a unit trip. This
proposed one-time amendment will be superseded by Amendment No. 208
when it is implemented.
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 proposed one-time change would extend 18-month instrument
surveillance intervals by a maximum of 39 days to March 31, 1996,
for specific Reactor Protective System (RPS), Engineered Safety
Features Actuation System (ESFAS), Power-Operated Relief Valve, Low
Temperature Overpressure Protection (LTOP), Remote Shutdown, Post-
Accident Monitoring (PAM), Radiation Monitoring, and Containment
Sump Level instruments.
The purpose of the RPS is to effect a rapid reactor shutdown if
any one or a combination of conditions deviates from a pre-selected
operating range. The system functions to protect the core and the
Reactor Coolant System (RCS) pressure boundary. The purpose of the
ESFAS is to actuate equipment which protects the public and plant
personnel from the accidental release of radioactive fission
products if an accident occurs, including a loss-of-coolant
accident, main steam line break, or loss of feedwater event. The
safety features function to localize, control, mitigate, and
terminate such incidents in order to minimize radiation exposure to
the general public. The PAM instruments provide the Control Room
operators with primary information necessary to take manual actions,
as necessary, in response to design basis events, and to verify
proper system response to plant conditions and operator actions. The
purpose of the Remote Shutdown System is to provide plant parameter
indications to operators on a Remote Shutdown Panel to be used while
placing and maintaining the plant in a safe shutdown condition in
the event the Control Room is uninhabitable. The indications are
used to verify proper system response to plant conditions and
operator actions. The LTOP System protects against RCS
overpressurization at low temperatures
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by a combination of administrative controls and hardware. Power-
Operated Relief Valves are set to lift before pressurizer safety
valves, and subsequently reseat to minimize the release of reactor
coolant from the RCS. The Containment Sump High Level Alarm System
provides an alarm in the Control Room to provide one of the
available indications of excessive RCS leakage during normal plant
operation. The Containment Area High Range Radiation Monitoring
System provides an indication of high radiation levels in
containment.
Failure of any of these systems is not an initiator for any
previously evaluated accident. Therefore, the proposed change would
not involve an increase in the probability of an accident previously
evaluated.
Surveillance and maintenance history has demonstrated good
capability for identifying adverse operation by individual
instruments. Baltimore Gas and Electric Company has the capability
to respond to an inoperable instrument by following the Technical
Specification Actions for an inoperable instrument or by performing
a channel calibration with the Unit at full power. However,
calibration of all the instruments at power is not desirable because
of personnel safety, personnel radiation protection goals, and plant
reliability concerns.
These factors provide assurance that the requested surveillance
extension will not adversely affect our ability to detect
degradation of the instruments. Also, either analysis is available
to show the instruments will operate properly during the requested
surveillance extension, or the surveillance program has shown that
problems will be identified and addressed appropriately. Therefore,
these channels will be able to perform the functions assumed in the
safety analysis, and there is no significant increase in the
consequences of an accident previously evaluated.
Therefore, the proposed Technical Specification changes do not
significantly 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 increase in surveillance interval for RPS, ESFAS,
Power-Operated Relief Valve, LTOP, Remote Shutdown, PAM, Radiation
Monitoring, and Containment Sump Level instrument surveillances does
not involve a significant change in the design or operation of the
plant. No plant hardware is being modified as part of the proposed
change. The proposed change also does not involve any new or unusual
actions by plant operators. 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 RPS, ESFAS, Power-Operated Relief Valve, LTOP, Remote
Shutdown, PAM, Radiation Monitoring, and Containment Sump Level
instruments are designed to provide actuation signals and/or
indications to ensure appropriate action is taken in response to
design basis accidents. Channel checks, channel functional tests and
routine comparison of the redundant and independent parameter
indications provides a reliable indication of instrument operation.
Also, either analysis is available to show the instruments will
operate properly during the requested surveillance extension, or
instrument surveillance program has shown that problems will be
identified and addressed appropriately. During the requested
extension, these systems will be available to perform the functions
assumed in the Safety Analysis. Surveillance and maintenance history
have demonstrated good capability for identifying adverse operation
by individual instruments. Baltimore Gas and Electric Company has
the capability to respond to such adverse operation, including
performing channel calibrations at power. However, such work on all
the instruments is not desirable because of personnel safety,
personnel radiation protection goals, and plant reliability
concerns. Extending the surveillance interval provides additional
possibility for instrument components to malfunction by means such
as drift or instrument failure, which could allow plant parameters
to exceed design bases assumptions. We have determined that the
effect of the surveillance interval extension on safety is small,
and operation of the instruments in the extended interval would not
invalidate any assumption in the plant licensing basis.
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: 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: Ledyard B. Marsh.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Units 1 and 2, Ogle County, Illinois, Docket Nos. STN
50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County,
Illinois
Date of amendment request: October 3, 1995.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) for both stations to
implement 10 of the line item TS improvements recommended in Generic
Letter (GL) 93-05, ``Line-Item Technical Specifications Improvements to
Reduce Surveillance Requirements for Testing During Power Operation,''
dated September 27, 1993. The proposed changes also include editorial
changes on the affected TS pages.
The proposed changes from GL 93-05 are the following: (1) TS
4.1.3.1.2 (GL 93-05, Item 4.2), extending the interval for checking the
operability of each full-length rod not fully inserted in the core from
31 days to 92 days; (2) Table 4.3-3 (GL 93-05, Item 5.14), extending
the interval for the digital channel operational test for radiation
monitoring instrumentation in the table from monthly to quarterly; (3)
TS 4.4.3.2 (GL 93-05, Item 6.6), extending the interval between current
tests of the required groups of pressurizer heaters from 92 days to
each refueling outage; (4) TS 4.4.6.2.2.b (GL 93-05, Item 6.1),
extending the time the plant may be in cold shutdown before pressure
isolation valve testing is required, prior to entry into Operational
Mode 2, from 72 hours to 7 days; (5) TS 4.5.1.1.b (GL 93-05, Item 7.1),
revising the requirement to verify the boron concentration in an
accumulator within 6 hours of any volume increase to the accumulator
(greater than or equal to 70 gallons) so that the verification is not
required when the volume increase is from the refueling water storage
tank (RWST) and the RWST has not been diluted since verifying that the
boron concentration of the RWST is within the concentration limits for
the accumulators; (6) TS 4.6.2.1 (GL 93-05, Item 8.1), extending the
interval between tests to verify each containment spray nozzle is
unobstructed from 5 years to 10 years; (7) TS 4.6.4.1 (GL 93-05, Item
5.4), extending the interval for testing each hydrogen monitor for
combustible gas control from 31 days to 92 days for the analog channel
operational test, and from 92 days to each refueling outage for channel
calibration; (8) TS 4.6.4.2 (GL 93-05, Item 8.5), extending the
interval between tests to demonstrate operability of the hydrogen
recombiner system from 6 months to once each refueling outage; (9) TS
4.7.1.2.1.a (GL 93-05, Item 9.1), extending the interval between tests
of the auxiliary feedwater pumps from 31 days to 92 days on a staggered
test basis; and (10) TS 4.11.2.6 (GL 93-05, Item 13), extending the
interval for determining the quantity of radioactivity contained in
each gas decay tank, when radioactivity is being added to the tanks,
from 24 hours to 7 days, with the 24-hour frequency maintained during
the primary coolant degassing operation. The editorial changes are the
following: (1) TS 4.4.6.2.1.c, changes the word ``from'' to the word
``to,'' (2) TS 4.5.1.1.c, the
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change clarifies that the motor control center compartment is for each
accumulator isolation valve, (3) TS 4.5.1.2, deletes the footnote
because the operating cycle in the footnote is over for each unit, and
(4) TS 4.7.1.2.1.a.2 and 4.7.1.2.1.c, renumbers and rephrases (only TS
4.7.1.2.1.a.2) other surveillance requirements for the auxiliary
feedwater pumps because of the proposed change to TS 4.7.1.2.1.a to
implement GL 93-05, Item 9.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The changes are consistent with GL 93-05 and NUREG-1366
[''Improvements to Technical Specifications Surveillance
Requirements,'' December 1992. In GL 93-05, the staff stated that it
concluded, in performing the study documented in NUREG-1366, that
safety can be improved, equipment degradation decreased, and an
unnecessary burden on licensee personnel eliminated by reducing the
frequency of certain testing required in the Technical
Specifications during power operation]. The changes eliminate
testing that is likely to cause transients or excessive wear of
equipment. An evaluation of these changes indicates that there will
be a benefit to plant safety. The evaluation, documented in NUREG-
1366, considered (1) unavailability of safety equipment due to
testing, (2) initiation of significant transients due to testing,
(3) actuation of engineered safety features that unnecessarily cycle
safety equipment, (4) importance to safety of that system or
component, (5) failure rate of that system or component, and (6)
effectiveness of the test in discovering the failure.
As a result of the decrease in the testing frequencies, the risk
of testing causing a transient and equipment degradation will be
decreased, and the reliability of the equipment will not be
significantly decreased.
The initial conditions and methodologies used in the accident
analyses remain unchanged. The proposed changes do not change or
alter the design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, accident
analyses results are not impacted. Appropriate testing will continue
to assure that equipment and systems will be capable of performing
the intended function.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes either modify allowable intervals between
certain surveillance tests, delete surveillance requirements, or
alter an action statement with regard to the required testing. The
proposed changes do not affect the design or operation of any
system, structure, or component in the plant. The safety functions
of the related structures, systems, or components are not changed in
any manner, nor is the reliability of any structure, system, or
component reduced by the revised surveillance or testing
requirements.
Appropriate testing will continue to assure that the system is
capable of performing its intended function. The changes do not
affect the manner by which the facility is operated and do not
change any facility design feature, structure, system, or component.
No new or different type of equipment will be installed. Since there
is no change to the facility or operating procedures, and the safety
functions and reliability of structures, systems, or components are
not affected, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
All of the proposed technical specification changes are
compatible with plant operating experience and are consistent with
the guidance provided in GL 93-05 and NUREG-1366. The changes
eliminate unnecessary testing that increases the risk of transients
and equipment degradation. There is no impact on safety limits or
limiting safety system settings.
The remaining proposed changes are administrative in nature and
have no impact on the margin of safety of any technical
specification. They do not affect any plant safety parameters or
setpoints.
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: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-373, LaSalle County
Station, Units 1, LaSalle County, Illinois
Date of amendment request: October 2, 1995
Description of amendment request: The proposed amendments would
revise Section 3.4.2 to change the safety/relief valve (SRV) safety
function lift setting tolerances from +1%, -3% to plus or minus 3% and
include as-left SRV safety function lift setting tolerances of plus or
minus 1%.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The staff has reviewed the licensee's analysis against
the standards of 10 CFR 50.92(c). The NRC staff's review is presented
below.
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The probability of an accident previously evaluated will not
increase as a result of this change, because the only changes are the
tolerances for the SRV opening setpoints and the speed of the reactor
core isolation cooling system (RCIC) turbine and pump. Changing the
maximum allowable opening setpoint for the SRVs does not cause any
accident previously evaluated to occur, or degrade valve or system
performance in any way so as to cause an accident to occur with an
increased frequency. In addition, the increased speed of the RCIC
turbine and pump are within the design limits of the system. RCIC
operability and failure probabilities are not impacted by this change.
The consequences of an ASME Overpressurization Event are not
significantly increased and do not exceed the previously accepted
licensing criteria for this event. General Electric (GE) has calculated
the revised peak vessel pressure for LaSalle Station to be 1341 psig,
which is well below the 1375 psig criterion of the ASME Code for upset
conditions, referenced in Section 5.2.2, Overpressurization Protection,
of the Updated Final Safety Analysis Report (UFSAR), and NUREG-0519
(Safety Evaluation Report related to the operation of LaSalle County
Station, Units 1 and 2, March 1981), and Section 15.2-4, Closure of
Main Steam Isolation Valves (BWR) of NUREG-0800 (Standard Review Plan).
GE has also performed an analysis of the limiting Anticipated
Transient Without Scram (ATWS) event, which is the Main Steam Isolation
Valve (MSIV) Closure Event. This analysis calculated the peak vessel
pressure to be 1457 psig, which is sufficiently below the 1500 psig
criterion of the ASME Code for emergency conditions.
Per NUREG-0519, listed above, Section 5.4.1, and Technical
Specification 4.7.3.b, the RCIC pump is required to develop flow
greater than or
[[Page 58399]]
equal to 600 gpm in the test flow path with a system head corresponding
to reactor vessel operating pressure when steam is supplied to the
turbine at 1000 +20, -80 psig. Increasing the turbine and pump speed
ensures these criteria will still be met and the consequences of an
accident will not increase.
Therefore, there is not a significant increase in the consequences
of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The only physical changes are to increase the allowable tolerances
for SRV opening setpoints and to increase the RCIC pump and turbine
speeds. These changes do not result in any changed component
interactions. The SRVs and RCIC will still provide the functions for
which they were designed. Since all of the other systems evaluated will
continue to function as intended, the proposed changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction in
the margin of safety.
While the calculated peak vessel pressures for the ASME
Overpressurization Event and the MSIV closure ATWS Event are larger
than that previously calculated without the proposed setpoint tolerance
increases, the new peak pressures remain sufficiently below the
respective licensing acceptance limits associated with these events. In
addition, the actual L1C8 reload analysis of the ASME
Overpressurization Event will be verified to be within the licensing
acceptance limit for that event prior to Unit 1 Cycle 8 startup, as
required in the normal reload 10 CFR 50.59 process. These licensing
acceptance limits have been previously evaluated as providing a
sufficient margin of safety. For other accidents and transients, the
increased setpoint tolerances have a negligible effect on the results,
so the margin of safety is preserved.
The staff has reviewed the amendment request and the licensee's no
significant hazards consideration determination. Based on the review
and the above discussions, the staff proposes to determine that the
proposed changes do not involve a significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: October 17, 1995.
Description of amendment request: The proposed amendment would
modify the Palisades Facility Operating License to reference 10 CFR
Part 40, allow the use of source materials as reactor fuel, delete
references to specific amendments and specific revisions in the listed
titles of the Physical Security Plan Suitability Training and
Qualification Plan and the Safeguards Contingency Plan, delete
paragraph 2.F on reporting requirements, and make minor editorial
changes. In addition, the Technical Specifications (TS) would be
modified as follows: (1) TS 3.1.2 would be modified to change the
pressurizer cooldown limit from 100 deg.F to 200 deg.F/hour; (2) the
shield cooling system requirements would be relocated to the Palisades
Final Safety Analysis Report (FSAR); (3) several minor editorial
changes to various sections of the TS are proposed; and (4) revisions
to several TS bases pages are proposed.
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:
Administrative Changes
Since these changes have no effect on the physical plant or its
operation, they cannot involve a significant increase in the
probability or consequences of an accident previously evaluated,
create the possibility of a new or different kind of accident from
any previously evaluated, or involve a significant reduction in a
margin of safety.
Technical Changes
The following evaluation supports the finding that operation of
the facility in accordance with the two non-administrative changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Use of Source Material as reactor fuel: The use of depleted or
natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in
addition to the currently allowed ``slightly enriched uranium''
would not affect the physical plant or its operation in any way
which could increase the probability of any previously evaluated
accident. Its use would not introduce any new kind or additional
amount of fission product material. Therefore, use of source
material as reactor fuel would not affect the consequences of an
accident previously evaluated.
Restoration of the Pressurizer Cooldown Rate Limit: The
Palisades Technical Specifications contain a single limit, item
3.1.2 b, for both heatup and cooldown rates for the pressurizer. The
October 5, 1994 change request proposed changing that limit from
200 deg.F/hour to 100 deg.F/hour solely due to its inconsistency
with the pressurizer design analysis. Fatigue calculations in the
pressurizer design analysis assumed a heatup rate of 100 deg.F/hour
and a cooldown rate of 200 deg.F/hour. Until issuance of Amendment
163, the Technical specifications contained a single limit for both
heatup and cooldown rates of 200 deg.F/hour. Although the installed
equipment is not capable of exceeding the 100 deg.F/hour heatup
limit, the October 5, 1994 change request proposed a revised limit
to assure that the Technical Specification limit was not less
restrictive than the design analysis. The higher pressurizer
cooldown rate does not affect the results of our analyses which
determined the PCS Pressure-Temperature limits or the [Loss of
Temperature Overpressurization] LTOP setting requirements of the
Technical Specifications.
When the change was proposed, it was not realized that the more
limiting cooldown rate might adversely, and unnecessarily, affect
plant operation. This proposed change to the Technical
Specifications would separate the limits for heatup rate and
cooldown rate, returning the specified cooldown rate to the original
value which was consistent with plant design. The current heatup
rate limit, which is also consistent with the design, would be
retained. The proposed pressurizer cooldown rate will allow
depressurizing of the primary coolant system [PCS] and flooding the
pressurizer steam space without undue restriction. The more rapid
depressurization would be important in the event of a steam
generator tube rupture.
Therefore, operation of the facility in accordance with the
proposed change to the Technical Specifications would 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 previously evaluated.
Use of Source Material as reactor fuel: The use of depleted or
natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in
addition to the currently allowed ``slightly enriched uranium''
would not affect the design (other than the fuel enrichment),
configuration, or operation of the plant. Therefore this change
cannot create the possibility of a new or different kind of accident
from any previously evaluated.
Restoration of the Pressurizer Cooldown Rate Limit: The proposed
change to the Technical Specifications would bring the plant within
the assumptions of the design documents for the pressurizer and in
line with the Accident analysis for the rapid reduction of the
primary coolant system pressure. With the lower rate specified in
the present technical specification, the depressurization of the PCS
will be delayed to maintain the lower pressurizer cooldown rate.
Therefore, operation of the facility in accordance with the
proposed change to the
[[Page 58400]]
Technical Specifications would not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
Use of Source Material as reactor fuel: The use of depleted or
natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in
addition to the currently allowed ``slightly enriched uranium''
would not affect the Safety Limits, Limiting Conditions for
Operation or other operating limits, or the safety analyses which
they support. Therefore, the margin of safety is unaffected.
Restoration of the Pressurizer Cooldown Rate Limit: The proposed
change to the Technical Specifications would bring the plant in line
with the design analysis. This will not reduce the margin of safety
since the higher rate is the basis for the present margin of safety.
Therefore, the proposed change to the Technical Specifications
would 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: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: Brian E. Holian, Acting.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: September 20, 1995.
Description of amendment request: The proposed amendment would
allow a one-time extension of the 18-month surveillance intervals
contained in the Technical Specifications (TS) related to system
testing, instrumentation calibration, component inspection, component
testing, response time testing and logic system functional tests for
various systems, components and instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS changes involve a one-time only change in the
surveillance testing intervals to facilitate a one-time only change
in the Fermi 2 operating cycle. The proposed TS changes do not
physically impact the plant nor do they impact any design or
functional requirements of the associated systems. That is, the
proposed TS changes do not significantly degrade the performance or
increase the challenges of any safety systems assumed to function in
the accident analysis. The proposed TS changes affect only the
frequency of the surveillance requirements and do not impact the TS
surveillance requirements themselves. In addition, the proposed TS
changes do not introduce any new accident initiators since no
accidents previously evaluated have as their initiators anything
related to the change in the frequency of surveillance testing.
Also, the proposed TS changes do not significantly affect the
availability of equipment or systems required to mitigate the
consequences of an accident because of other, more frequent testing
or the availability of redundant systems or equipment. Furthermore,
a historical review of surveillance test results support the above
conclusions. Therefore, the proposed TS changes do not significantly
increase the probability or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes involve a one-time only change in the
surveillance testing intervals to facilitate the one-time only
change in the Fermi 2 operating cycle. The propose TS changes do not
introduce any failure mechanisms of a different type than those
previously evaluated since there are no physical changes being made
to the facility. In addition, the surveillance test requirements
themselves will remain unchanged. 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.
Although the proposed TS changes will result in an increase in
the interval between some surveillance tests, the impact, if any, on
system availability is small based on other, more frequent testing
or redundant systems or equipment, and there is no evidence of any
time dependent failures that would impact the availability of the
systems. Therefore, the assumptions in the licensing basis are not
impacted, and the proposed TS changes do not significantly reduce 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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Brian E. Holian, Acting.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: August 8, 1995.
Description of amendment request: The amendments would revise
Technical Specification Section 3/4.4.8, Table 4.4-4, Table Notations,
to allow the reactor coolant system gross specific activity measurement
method to be changed from the current degassed method to a non-
degassed, or pressurized dilution, method.
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
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The amendments will have no effect on the probability of
the occurrence of any accident. It has been demonstrated that the
results obtained by the pressurized dilution technique are
statistically similar to results obtained by the degassed technique.
Therefore, implemention of the new method will have no effect
insofar as the accuracy of the NC [reactor coolant system] system
specific activity determination is concerned. Therefore, there will
be no effect upon any accident dose consequences.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No accident causal mechanisms will be affected by
installation of the sampling equipment required by the pressurized
dilution technique. Operation of the NC system itself will not be
affected by the proposed change in sampling technique. All procedure
changes required for implementation of the new sampling method will
be made according to the provisions of 10 CFR 50.59. No impact on
other areas of plant operations will be generated as a result of the
new sampling method.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. No impact on any safety limits will
result from the change in sample method from the degassed technique
to the pressurized dilution technique. Several benefits will result
from the change,
[[Page 58401]]
including fewer opportunities for valve mispositionings to occur, as
well as reduced radiation exposure to Chemistry technicians. The
proposed amendment is consistent with a similar amendment approved
by the NRC for McGuire Nuclear Station (Amendment Nos. 66 and 47 for
McGuire Units 1 and 2, respectively).
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 7, 1995.
Description of amendment request: The proposed change would revise
Technical Specification 3/4.5.1 SAFETY INJECTION TANKS (SITs) by
increasing the specified range associated with SIT water level and
nitrogen cover pressure.
The current limiting conditions for operation (LCO) for the SIT
requires that four SITs be operable with a water volume in the range of
1679 cubic feet (78%) to 1807 cubic feet (83.8%) and a nitrogen cover
pressure between 600 psig to 625 psig. The proposed change requests an
expanded range of 925.6 cubic feet (40%) to 1807 cubic feet (83.8%) for
SIT level and 600 psig to 670 psig for SIT pressure indicators.
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 facility in accordance with this change does
not involve an increase in the probability of any accident. The SITs
are used to mitigate the consequences of an accident and are not
accident initiators.
The proposed change would actually decrease the consequence of
events such as LOCA [loss of coolant accident] which would result in
rapid RCS [reactor coolant system] depressurization.
By reducing SIT level, the initial nitrogen gas volume is
increased which results in an increase in the SIT flow rate into the
RCS for a given RCS pressure transient. This decreases the time
required to fill the reactor vessel lower plenum after the end of
blowdown. During refill, fuel cladding temperature increases rapidly
due to insufficient cooling which is provided solely by rod to rod
thermal radiation. Decreasing the refill time therefore, results in
lower cladding temperature at the start of core reflood which
results in lower Peak Cladding Temperature (PCT) during reflood.
Increasing the nitrogen cover pressure would also result in
increased SIT flow rate and would be beneficial as described above.
Therefore, the proposed change will not involve a significant
increase in the probability or consequence of any accident.
The proposed change will not create any new system connections
or interactions. Thus, no new modes of failure are introduced. The
increased range for SIT pressure and level is actually beneficial in
maintaining lower PCT following a LOCA.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The impact of the proposed changes on the Waterford 3 FSAR
[Final Safety Analysis Report] analyses have been evaluated. The AOR
[Analysis of Record] shows that PCT and maximum cladding oxidation
would increase slightly as a result of this change. However, they
both remain below the acceptance criteria values of 2200 degrees
fahrenhit and 17% for PCT and maximum cladding oxidation,
respectively. The system capabilities to mitigate the consequences
of accidents will be the same as they were prior to these changes.
Therefore, the proposed changes do[es] 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: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street NW, Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: August 10, 1995
Description of amendment request: This amendment would incorporate
certain improvements into the Three Mile Island, Unit 1 Technical
Specifications consistent with the Standard Technical Specifications
for Babcock and Wilcox plants. The requested changes would affect the
reactor building isolation instrumentation, sampling frequency for the
sodium hydroxide tank, and the surveillance requirements for the plant
vital bus batteries.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment involves changes to the
TMI-1 Technical Specifications [TS] which are consistent with the
[Babcock & Wilcox] B&W Standard Technical Specifications ([R]STS),
NUREG-1430. This change does not involve any change to system or
equipment configuration. The proposed amendment revises certain
surveillance requirements, or extends certain surveillance
intervals. The reliability of systems and components relied upon to
prevent or mitigate the consequences of accidents previously
evaluated is not degraded by the proposed changes. Therefore, this
change does not involve a significant increase in the probability of
occurrence or the 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 change
only involves changes to surveillance requirements that are
consistent with RSTS or deletion of requirements which are not
appropriate for TS. No new failure modes are created and thus the
changes are bounded by accidents previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. These proposed changes involve deletions of requirements or
changes in surveillance requirements consistent with the B&W RSTS.
No operating limits are affected and no reduction in the margin of
safety is involved.
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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
[[Page 58402]]
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Phillip F. McKee.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London, Connecticut
Date of amendment request: October 24, 1995.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) Surveillance Requirement of
Section 4.4.5.1, ``Steam Generators'' and the Bases for Section 3/
4.4.5, ``Steam Generators.'' Typographical errors in Section
4.4.5.1.3.c.1 and Table 4.4-6 are also proposed to be corrected. The
proposed amendment would defer the next required surveillance to
inspect steam generator tubes from October 20, 1996, to the next
refueling outage or no later than October 20, 1997, whichever is
earlier.
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 (SHC), which is presented below:
Pursuant to 10 CFR 50.92, NNECO [the licensee] has reviewed the
proposed one-time change to extend the maximum allowable inspection
interval for steam generator tubes from 24 months to 36 months.
NNECO concludes that these changes do not involve a significant
hazards consideration since the proposed change satisfies the
criteria in 10 CFR 50.92(c). That is, the proposed changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
This change involves one-time deferment of the eddy current
inspection of the steam generator tubes until the end of the next
refueling outage following the thirteenth fuel cycle, but no longer
than 12 months beyond the original due date for the inspection. The
steam generator tubes have only been exposed to one operating cycle
and are made of thermally treated Alloy 690, one of the most
corrosion resistant material currently used in recirculating steam
generators. Following the first full fuel cycle of operation, the
steam generator tube inspection found the tubes to be in excellent
condition (i.e., no repairs were required and there was no evidence
of an active degradation mechanism). Accordingly, no significant
tube degradation is expected by the end of the thirteenth fuel
cycle. Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
This one-time change, allowing the steam generator tubes to be
examined at the end of the refueling outage following Cycle 13 does
not alter the physical design, configuration, or method of operation
of the plant. The extension of the inspection interval is not
expected to result in significant steam generator tube degradation.
Therefore, the proposed change does 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.
Steam generator tube degradation occurs primarily during
operation. The change to extend the maximum allowable inspection
interval for steam generator tubes from 24 months to 36 months will
not significantly increase the total operating time during Cycle 13
(the plant was in an outage for at least 10 months of the 12 month
extension). Therefore, there is no significant effect on the extent
and severity of tube degradation. The improved corrosion resistance
of the steam generators tubes (thermally treated Alloy 690)
minimizes the threat of primary- and secondary-side corrosion. No
indications of corrosion have been identified in inspections
performed so far. Based on our assessment of the inspection data and
corrosion potential, all tubes are expected to be within the
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR Steam
Generator Tubes,'' limits by the end of Cycle 13. Also, correction
of the typographical errors will improve the fidelity of the
specification. 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: May 1, June 14 and 29, July 14, 17, 18,
and 26, 1995 with supplemental information provided by letter dated
October 20, 1995.
Description of amendment request: Each proposed amendment would
change the surveillance requirement frequency from the current once per
18-month interval to once per 24-month which is the current length of a
Millstone Unit 3 refueling cycle. The changes pertain to the following
equipment:
May 1, 1995, Flow Paths--Operating; Position Indication System; Rod
Drop Time; Seismic Monitoring System; Loose Part Detection System;
Quench Spray System; Containment Recirculation Spray System;
Containment Isolation Valves. This notice supersedes the notice
published in the Federal Register on June 6, 1995 (60 FR 29882)
relating to containment isolation valves.
May 1, 1995, Steam Generator Tube Inspections; 10CFR50, Appendix J,
Type B and Type C Tests.
June 14, 1995, AC Sources Operating; DC Sources Operating;
Containment Penetration Conductor Overcurrent Protective Devices;
Motor-Operated Valves Thermal Overload Protection.
June 29, 1995, Electric Hydrogen Recombiners; Auxiliary Feedwater
System; Reactor Plant Component Cooling Water System; Service Water
System; Snubbers.
July 14, 1995, ECCS Subsystems--Tavg Greater Than or Equal to 350
deg.F; pH Trisodium Phosphate Storage Baskets.
July 17, 1995, Supplementary Leak Collection and Release System;
Control Room Emergency Ventilation System; Control Room Envelope
Pressurization System; Auxiliary Building Filter System; Fuel Building
Exhaust Filter System.
July 18, 1995, Reactor Coolant System.
July 26, 1995; Reactor Trip System Instrumentation; ESFAS
Instrumentation; Remote Shutdown Instrumentation; Accident Monitoring
Instrumentation; RCS Total Flow Rate; Process and Radiation Monitoring
Instrumentation.
In addition, the specifications are changed from a five-column to a
one-column format.
Basis for proposed no significant hazards consideration
determination: The Commission has made a proposed determination that
the amendment request involves 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
[[Page 58403]]
(3) involve a significant reduction in a margin of safety. As required
by 10 CFR 50.91(a), the licensee has provided its analysis of the issue
of no significant hazards consideration. The NRC staff has reviewed the
licensee's analysis against the standards of 10 CFR 50.92(c). The NRC
staff's review is presented below:
1. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to surveillance requirements of the Millstone
Unit No. 3 Technical Specifications extend the frequency for checking
the operability of the affected components/equipment. The proposal
would extend the frequency from at least once per 18 months to at least
once each refueling interval (i.e., nominal 24-months).
Changing the frequency of surveillance requirements from at least
once per 18 months to at least once each refueling interval does not
change the basis for the frequency. The frequency was chosen because of
the need to perform this verification under the conditions that apply
during a plant outage, and to avoid the potential of an unplanned
transient if the surveillances were conducted with the plant at power.
The proposed changes do not alter the intent or method by which the
surveillances are conducted, do not involve any physical changes to the
plant, do not alter the way any structure, system, or component
functions, and do not modify the manner in which the plant is operated.
As such, the proposed changes in the frequency of surveillance
requirements will not degrade the ability of the equipment/components
to perform its safety function.
Additional assurance of the operability of the components/equipment
is provided by additional surveillance requirements (e.g., monthly or
quarterly surveillances).
Equipment performance over the last four operating cycles was
evaluated to determine the impact of extending the frequency of
surveillance requirements. This evaluation included a review of
surveillance results, preventive maintenance records, and the frequency
and type of corrective maintenance. It concluded that there is no
indication that the proposed extension could cause deterioration in the
condition or performance of any of the subject components.
In addition to the substantive changes, there are format changes
which are merely editorial and because format changes produce no
physical change they do not influence the probability or consequences
of accidents.
Since the proposed changes only affect the surveillance frequency
for safety systems that are used to mitigate accidents, the changes
cannot affect the probability of any previously analyzed accident.
While the proposed changes can lengthen the intervals between
surveillances, the increases in intervals has been evaluated and it is
concluded that there is no significant impact on the reliability or
availability of the safety system and consequently, there is no impact
on the consequences on any analyzed accident.
2. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes to surveillance requirements of the Millstone
Unit No. 3 Technical Specifications extend the frequency for verifying
the operability of the affected components/equipment. The proposal
would extend the frequency from at least once per 18 months to at least
once each refueling interval (nominal 24 months).
Changing the frequency of surveillance requirements from at least
once per 18 months to at least once each refueling interval does not
change the basis for the frequency. The frequency was chosen because of
the need to perform this verification under the conditions that apply
during a plant outage, and to avoid the potential of an unplanned
transient if the surveillances were conducted with the plant at power.
In addition to the substantive changes, there are format changes
which are merely editorial and because format changes produce no
physical change they do not influence the probability of new or
different types of accidents.
The proposed changes do not alter the intent or method by which the
surveillances are conducted, do not involve any physical changes to the
plant, do not alter the way any structure, system, or component
functions, and do not modify the manner in which the plant is operated.
As such, the proposed changes cannot create the possibility of a new or
different kind of accident from any previously evaluated.
3. The changes do not involve a significant reduction in a margin
of safety.
The proposed changes to surveillance requirements of the Millstone
Unit No. 3 Technical Specifications extend the frequency for verifying
the operability of the components/equipment. The proposal would extend
the frequency from at least once per 18-months to at least once each
refueling interval (24-months).
In addition to the substantive changes, there are format changes
which are merely editorial and because format changes produce no
physical change they do not influence the margin of safety.
The proposed changes to surveillance frequency are still consistent
with the basis for the frequency, and the intent or method of
performing the surveillance is unchanged. Further, the current
inservice testing requirements and the previous history of reliability
of the system provides assurance that the changes will not affect the
reliability of the auxiliary feedwater system. Thus, it is concluded
that there is no impact on the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: September 29, 1995.
Description of amendment requests: The amendments would add a one-
time footnote to the Technical Specifications regarding the emergency
diesel generator diesel fuel oil storage and transfer system to permit
the existing storage tanks to be replaced with double walled tanks and
piping that comply with new California regulations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Neither the emergency diesel generators (EDGs) nor the diesel
fuel oil (DFO) storage and transfer system is an accident initiator.
When performing the modifications to the
[[Page 58404]]
DFO storage tanks and transfer piping, administrative compensatory
measures will be taken to reduce the potential challenge to the EDGs
and to verify the operability of the DFO transfer system. A
probabilistic risk assessment (PRA) was performed and demonstrates
that the change in core damage frequency associated with taking each
DFO storage tank and its associated suction transfer piping out of
service for 60 days (total of 120 days for both trains) is not
significant considering the compensatory measures which will be
taken during the tank replacement period.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Neither the EDGs nor the DFO storage and transfer system is an
accident initiator. Temporary DFO storage will be onsite during tank
replacement. The fire protection guidelines in Appendix 9.5B of the
Updated Final Safety Analysis Report will be complied with in order
to ensure temporary DFO storage without risk to plant systems.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes considering implementation of the
compensatory measures has been shown to not impair safe operation of
the plant. Having one DFO storage tank and associated piping out of
service does not reduce the margin of safety since temporary storage
of DFO will be maintained onsite and administrative compensatory
measures will be taken to minimize the potential impact of this
condition. Additionally, delivery of DFO to the site is available
within 24 hours if needed.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: October 4, 1995.
Description of amendment requests: The amendments would relocate
the requirements in ten sub-sections of the Technical Specifications to
licensee controlled documents in accordance with the guidance in the
Commission's Final Policy Statement and the Commission's revisions to
10 CFR 50.36 (60 FR 36959, July 19, 1995) on the content of Technical
Specifications and the Standard Technical Specifications, Westinghouse
Plants, NUREG-1431, Rev. 1, dated April 1995. The ten sub-sections
which the licensee proposes to relocate, without changes to the
requirements, to the Updated Final Safety Analysis Report or other
controlled documents relate to: boration system flow path, position
indication system, rod drop time, seismic instrumentation, chlorine
detection system, turbine overspeed protection, containment leakage,
containment structural integrity, electrical equipment protective
devices and containment penetration conductor overcurrent protective
devices.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes simplify the Technical Specifications (TS),
meet regulatory requirements for relocated TS, and implement the
recommendations of the Commission's Final Policy Statement on TS
Improvements and revised 10 CFR 50.36. Future changes to these
requirements will be controlled by 10 CFR 50.59. The proposed
changes are administrative in nature and do not involve any
modifications to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety-related system performs
its function. Also, no changes to the operation of the plant or
equipment are involved.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes involve relocating TS requirements to a
licensee-controlled document. The requirements to be relocated were
identified by applying the criteria endorsed in the Commission's
Final Policy Statement, which is included in the new revision of 10
CFR 50.36, and are consistent with NUREG-1431, Rev. 1 (Reference 2).
Thus, the proposed changes do not alter the basic regulatory
requirements and do not affect any safety analysis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of amendment request: November 2, 1995.
Description of amendment request: The proposed amendment would
revise Section 5.0, Administrative Controls, of the Trojan Nuclear
Plant Technical Specifications, Appendix A to License NPF-1, to reflect
changes in the organization of the Portland General Electric Company
(PGE) as they apply to oversite and management of the Trojan Nuclear
Plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
1. The requested license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The changes in management titles and reporting relationships are
administrative in nature, do not alter the intent of the Possession
Only License, and do not modify
[[Page 58405]]
the present plant systems or adminstrative controls necessary to
preserve and protect the integrity of the nuclear fuel at the Trojan
Nuclear Plant. The Trojan Site Executive and Plant General Manager
will be located at the site and will continue to provide senior
management attention to each of the functional areas in the Trojan
Nuclear Plant organization during decommissioning of the facility.
The general classification of accidents for the permanently
defueled condition are limited. The three classifications are (1)
radioactive release from a subsystem or component, (2) fuel handling
accident, and (3) loss of spent fuel decay heat removal capability.
The probability of occurrences of consequences from these accidents
remain unchanged and are bounded by the current accident analysis.
Therefore, the requested changes do not involve a significant
increase in the probability or occurrence of an accident previously
evaluated.
2. The requested license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The requested amendment is administrative in nature, does not
affect the manner in which systems and components are operated or
maintained, and does not alter the intent of the Possession Only
License. The accident scenarios associated with the permanently
defueled condition are limited to (1) radioactive release from a
subsystem or component, (2) fuel handling accident and (3) loss of
spent fuel decay heat removal capability. There are no new accident
scenarios or failure modes created by the requested administrative
changes. Therefore the requested change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The requested license amendment does not involve a
significant reduction in a margin of safety.
The requested amendment is administrative in nature, does not
affect the manner in which systems and components are operated or
maintained, does not alter the intent of the Possession Only
License, nor does it adversely impact previously accepted margins of
safety. Therefore, the requested amendment does not involve a
significant reduction in margin of safety.
The NRC staff has reviewed the analysis of the licensee 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: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensees: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRR Project Director: Seymour H. Weiss.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: October 7, 1995 as supplemented by
letter dated October 27, 1995.
Description of amendment request: The proposed change to Hope Creek
Technical Specifications (TSs) 4.8.1.1.2, ``A.C. Sources--Operating'',
would replace the reference to a voltage and frequency band for the 10
second starting time test with a minimum required voltage and frequency
that must be attained within 10 seconds.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident [* * *] previously evaluated.
Since no change is being made to the offsite power supplies, or
to any system or component that interfaces with the offsite power
supplies, there is no change in the probability of a Loss of Offsite
Power Accident.
Since the proposed change still ensures the surveillance
requirements meet the licensing basis and since the full spectrum of
loading, unloading and standby testing performed at the 18 month
frequency continues to demonstrate the capability of the diesel
generators to satisfy onsite power requirements during simulated
accident conditions while the monthly testing demonstrates
availability, there is no change in the consequences of an accident.
Since the proposed change will eliminate unnecessary adjustments
to the governor controls, the probability of malfunction is
potentially reduced.
This change ensures the surveillance requirements reflect the
design basis and provide a basis for consistent timing methodology.
Since the proposed change is consistent with the intent of the
existing specifications, and with the design basis of the system and
since no physical changes are being proposed, no action will occur
that will increase the probability or consequences of an accident or
malfunction of equipment important to safety. The diesel generators
will continue to function as stated in the UFSAR [Updated Final
Safety Analysis Report].
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident or
malfunction of equipment important to safety previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change does not result in any design or physical
configuration changes to the offsite power supplies or to the diesel
generators. Operation in accordance with the proposed change will
not impair the diesel generators ability to perform as provided in
the design basis. By eliminating unnecessary adjustments to the
diesel generator governor control, performance during any accident
is potentially enhanced. The diesel generators will continue to
function as stated in the UFSAR. Therefore, the proposed change will
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
Since the proposed change does not involve the addition or
modification of plant equipment, is consistent with the intent of
the existing Technical Specifications, meets the intent of
applicable Regulatory Guides, and is consistent with the design
basis of the diesel generators and the UFSAR, no action will occur
that will 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: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Attorney for licensee: M.J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street NW., Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.
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 amendment request: September 29, 1995.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3/4.4.3, Safety Valves and Pilot
Operated Relief Valve--Operating, and associated Bases 3/4.4.2 and 3/
4.4.3, Safety Valves, to increase the lift setting of the pressurizer
code safety valves (PSVs) to [equal to or less than] 2575 psig, which
corresponds to a lift setting tolerance of +3% of the nominal lift
pressure. Increasing the upper bound of the lift setting tolerance of
the PSVs from +1% to +3% will allow normal surveillance testing of the
PSVs to be within +3% of the nominal lift setpoint of 2500 psig, which
is still acceptable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 58406]]
consideration, which is presented below:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No.
1 in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because increasing the PSV lift
tolerance from +1% to +3% only affects the as-found tolerance of the
PSVs. The initial setting tolerance will still be limited to +1%. No
hardware modification will be done to the valves which could affect
any accident initiators.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because increasing the PSV lift
tolerance from +1% to +3% does not affect the radiological releases
of any accident previously evaluated in the [Updated Safety Analysis
Report] USAR. This is not a hardware modification and the reactor
coolant pressure boundary integrity is unaffected.
2. Not create the possibility of a new kind of accident from any
previously evaluated because increasing the PSV lift tolerance from
+1% to +3% allows the PSVs to protect the reactor coolant pressure
boundary from overpressure transients. This change only affects the
allowable lift tolerance. The initial lift setting tolerance is
still less than +1%. This change does not modify the valve hardware
or alter the operation of the valves. The possibility of the valves
spuriously opening during power operation will not be changed. The
valve setpoint with a -3% lift tolerance is well above the normal
operating conditions and the [reactor coolant system] RCS high
pressure trip setpoint.
3. Not involve a significant reduction in a margin of safety
because at the +3% lift tolerance the RCS pressure and the reactor
thermal power are still within the USAR acceptance criteria for a
control rod withdrawal at low power. This change ensures the
Technical Specification lift setpoint tolerances are consistent with
the requirements given in the [American Society of Mechanical
Engineers] ASME Boiler and Pressure Vessel Code.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: June 21, 1994, as amended by letter
dated October 23, 1995.
Description of amendment request: The proposed amendment would
relocate the review and audit requirements of the On-site Review
Committee (ORC) and Nuclear Safety Review Board (NSRB) contained in TS
6.5.1, TS 6.5.2 and TS 6.5.3 to the Operational Quality Assurance
Manual (OQAM). In addition, the proposed amendment would delete
reference to the Manager, Nuclear Safety and Emergency Preparedness in
TS 6.2.3. A revision to the Index was proposed to reflect the
relocations. This amendment request was previously published in the
Federal Register on August 31, 1994 (59 FR 45036).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes are administrative and equivalent descriptions and
requirements for these oversight committees are contained in the
OQAM.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
These changes do not involve any physical alterations to the
plant. There is no new type of accident or malfunction created and
the method and manner of plant operation will not change. The
changes are administrative and equivalent descriptions and
requirements for these oversight committees are contained in the
OQAM.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety remains unaffected since no design change
is made and plant operation remains the same. The changes are
administrative and equivalent descriptions and requirements for
these oversight committees are contained in the OQAM.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: William H. Bateman.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: October 17, 1995.
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Unit No. 2 (NA-2). Specifically, the proposed change would reduce from
two to one the minimum number of steam generators (SGs) required to be
opened for inspection during the first refueling outage following an SG
replacement. TS surveillance requirements 4.4.5.0 through 4.4.5.5 for
inspection of the SG tubes ensure that the structural integrity of this
portion of the Reactor Coolant System will be maintained. Accordingly,
the purpose of TS 4.4.5.1 is to require periodic sample inspections of
SGs. The initial inspection after SG replacement combined with the
subsequent inservice inspections serve to provide reasonable assurance
of detection of structural degradation of the tubes. The proposed TS
change does not affect or change this basis. However, the requirement
that two SGs would be opened and inspected during the first refueling
outage after SG replacement is considered unnecessary.
The NA-2 SGs were replaced during the first quarter of 1995. The
purpose of SG replacement was to restore the integrity of the SG tubes
to a level equivalent to new SGs. In reality, replacement SG components
incorporate a large number of design improvements which reflect the
``state-of-the-art'' technology that currently exists for SG design.
These design improvements will improve the long-term maintainability
and reliability of the replacement SGs. These enhancements do not
adversely affect the mechanical or thermal-hydraulic performance of the
SGs. Thus, the replacement SGs are considered superior to the original
SGs in terms of design and materials.
The proposed TS change does not affect or change any limiting
conditions for operation (LCO) or any other surveillance requirements
in the TS and the Basis for the surveillance requirement remains
unchanged. An inspection of the minimum required number of tubes will
still be performed
[[Page 58407]]
prior to returning the SGs to service. Although the proposed change
reduces the number of SGs required to be opened for inspection, the
minimum number of tubes required to be examined during the inspection
is not being changed. Thus, the minimum inspected tube population size
would not be changed.
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:
We have evaluated the proposed change against the criteria
described in 10 CFR 50.92 and concluded that the proposed Technical
Specifications change does not pose a significant hazards
consideration.
[1] The proposed Technical Specifications change does not affect
the assumptions, design parameters, or results of any UFSAR [Updated
Final Safety Analysis Report] accident analysis and the proposed
amendment does not add or modify any existing equipment. Therefore,
the proposed Technical Specifications change would not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
[2] The proposed change to the Technical Specifications does not
involve modifications to any of the existing equipment or affect the
operation of any existing systems. The absence of any hardware or
software changes means that the accident initiators remain
unaffected, so no unique accident possibility is created. Therefore,
the proposed Technical Specifications change would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
[3] Although the proposed change will reduce the minimum number
of steam generators required to be opened for inspection during the
first refueling outage following steam generator replacement, the
revised Technical Specification surveillance will continue to ensure
that a sampling of steam generator tubes will be inspected. The
operability of the steam generators will also continue to be
verified by periodic inservice inspections. Therefore, since
equipment reliability will be maintained, the proposed Technical
Specifications 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: David B. Matthews.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: October 18, 1995.
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TS) 3.4, ``Steam and Power Conversion System,'' by modifying and
clarifying the operability requirements for the main steam safety
valves (MSSVs), the auxiliary feedwater (AFW) System, and the
condensate storage tank system.
The proposed amendment would eliminate inconsistencies within TS
Section 3.4 and provide the basis for acceptable operation of the
Auxiliary Feedwater System below 15% reactor power. The proposed
amendment supersedes in its entirety a previously submitted proposed
amendment dated May 20, 1994, which was noticed in the Federal Register
on September 28, 1994 (59 FR 49442).
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:
Significant Hazards Determination for Proposed Changes to Technical
Specification (TS) 3.4.a ``Main Steam Safety Valves''
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Currently, TS 3.4.a.1.A.2 requires five MSSVs to be operable
prior to heating the reactor > 350 deg.F. The proposed change
requires a minimum of two MSSVs per steam generator to be operable
prior to heating the reactor coolant system > 350 deg.F, and five
MSSVs per steam generator to be operable prior to reactor
criticality. If these conditions cannot be met within 48 hours,
within 1 hour action shall be initiated to achieve hot standby
within 6 hours, achieve hot shutdown within the following 6 hours,
and achieve and maintain the reactor coolant system temperature <
350 deg.F within an additional 12 hours.
The MSSVs are relied upon to function in each of the following
USAR analyzed accidents: Reactor Coolant Pump Locked Rotor, Loss of
External Electrical Load, Loss of Normal Feedwater, Uncontrolled Rod
Cluster Control Assembly Withdrawal, Steam Generator Tube Rupture,
and Anticipated Transients without Scram.
In a subcritical condition, two operable MSSVs are capable of
relieving the maximum steam generated during these anticipated
design basis transient events. Because this proposed TS requires all
MSSVs to be operable prior to reactor criticality, there will be no
adverse effect on the health and safety of the public.
In all cases, the relieving capacity of the MSSVs is sufficient
to maintain steam pressures within safety analysis acceptable
criteria, and reactor criticality is not permitted unless all MSSVs
are operable. Therefore, there is no adverse effect on the health
and safety of the public and no significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not alter the plant configuration,
operating setpoints, or overall plant performance. Therefore, it
does not create the possibility of a new or different kind of
accident.
3. Involve a significant reduction in the margin of safety.
The USAR safety analysis assumes five MSSVs per steam generator
are operable. However, as shown above, this change results in no
steam generator overpressure event or increase in the radiological
dose. Therefore, this change will not involve a reduction in the
margin of safety.
Significant Hazards Determination for Proposed Changes to Technical
Specification (TS) 3.4.b ``Auxiliary Feedwater System''
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Current TS 3.4.a.1.A.1 and TS 3.4.b governing auxiliary
feedwater flow to the steam generators are being combined and
titled, ``Auxiliary Feedwater System.'' This change is consistent
with the format of ``Westinghouse Standard Technical
Specifications,'' NUREG-1431. In addition to the formatting changes,
a number of technical changes are being proposed. These are:
The correction of an inconsistency between current TS
3.4.a.1.A.1 and current TS 3.4.b.2.A.
The addition of a seven (7) day Limiting Condition for Operation
(LCO) action statement for one inoperable steam supply to the
turbine driven auxiliary feedwater pump.
A specification is being added to permit any of the following
conditions with reactor power less than 15%, without declaring the
corresponding AFW train inoperable: the AFW pump control switches
located in the control room to be in the ``pullout'' position, flow
control valves AFW-2A and AFW-2B to be in a throttled or closed
position, and train cross-connect valves AFW-10A and AFW-10B to be
in the closed position.
An inconsistency currently exists between current TS 3.4.a.1.A.1
and current TS
[[Page 58408]]
3.4.b.2.A. TS 3.4.a.1.A.1 requires the system piping and valves
directly associated with providing auxiliary feedwater flow to the
steam generators to be operable, with a corresponding 48 hour
limiting condition for operation (LCO) action statement if this
requirement is not met. TS 3.4.b.2.A allows one auxiliary feedwater
pump to be inoperable for 72 hours. This arrangement can cause a
conflict regarding which TS is applicable depending on which
component in the auxiliary feedwater flowpath to the steam
generators is inoperable. By moving all TS action statements to TS
3.4.b, the inconsistency between TS 3.4.a.1.A.1 and TS 3.4.b.2.A
will be eliminated. The requirement to maintain the operability of
the system piping and valves directly associated with providing
auxiliary feedwater flow to the steam generators remains, but is
being modified to prevent the removal of both AFW supply headers
from service.
Proposed TS 3.4.b.2.C is being added to allow one steam supply
to the turbine driven auxiliary feedwater pump to be inoperable for
seven days. This addition is consistent with ``Westinghouse Standard
Technical Specifications,'' NUREG-1431. The seven day completion
time is reasonable based on the redundant steam supplies to the
pump, the availability of the redundant motor-driven AFW pumps, and
the low probability of an event occurring that requires the
inoperable steam supply to the turbine driven AFW pump. For these
reasons, this change will have no adverse effect on the health and
safety of the public.
Proposed TS 3.4.b.6.A and B permit the AFW Pump control switches
located in the control room to be placed in the ``pull out''
position and valves AFW-2A and AFW-2B to be in a throttled position
when below 15% reactor power without declaring the corresponding AFW
train inoperable. This change is proposed to resolve concerns
regarding the cycling of the AFW pumps and the throttling of valves
AFW-2A and AFW-2B during plant startups and shutdowns. Analysis
shows that control room operators have a minimum of ten minutes to
initiate auxiliary feedwater flow after a design basis accident with
no steam generator dryout or core damage.
All accidents which rely on AFW flow for mitigation were
reanalyzed to support this change. These analyses were completed
assuming an initial power of 100%. However, a 15% reactor power
restriction has been imposed on placing the AFW pump control
switches located in the control room in the ``pull out'' position
and throttling valves AFW-2A and AFW-2B. This restriction in effect
limits use of TS 3.4.b.6 to plant startups, shutdowns and other low
power operating conditions.
This change alters the assumptions of the safety analysis for
the Small-Break Loss of Coolant Accident, the Steam Generator Tube
Rupture and the Loss of Normal Feedwater due to their dependence on
the AFW system to start and supply AFW for heat removal. To support
this change, the Westinghouse Electric Corporation performed an
analysis of the Small-Break Loss-of-Coolant Accident using the
NOTRUMP code assuming ten minutes for operator action to initiate
auxiliary feedwater. This analysis resulted in a Peak Cladding
Temperature (PCT) of 1053 deg.F from an initial power level of
100%. In addition, all other acceptance criteria of 10 CFR 50.46
were met. This large margin to the 2200 deg.F PCT limit supports
ten minutes for operator action to initiate auxiliary feedwater.
Furthermore, WPSC has analyzed the Loss of Normal Feedwater and
the Steam Generator Tube Rupture Accident assuming delays in the
initiation of auxiliary feedwater. The Loss of Normal Feedwater
Accident with a ten minute delay in the initiation of Auxiliary
Feedwater does not result in any adverse condition in the core. It
does not result in water relief from the pressurizer safety valves,
nor does it result in uncovering the tube sheets of the steam
generators. Also, at all times the Departure from Nucleate Boiling
Ratio (DNBR) remained greater than 1.30. The Steam Generator Tube
Rupture Accident with no auxiliary feedwater flow was also analyzed.
The results of this analysis indicate that neither steam generator
empties of liquid and at least 20 deg.F of reactor coolant system
subcooling is maintained throughout the transient. Also, there is no
increase in the radiological dose to the public.
Ten minutes is an acceptable time for operator action because
four independent alarms in the control room would initiate operator
action to place the AFW pump control switches to the ``auto''
position and initiate AFW flow to the steam generators when
necessary. These include two steam generator lo level alarms (one
per steam generator), and two steam generator lo-lo level alarms
(one per steam generator). Provisions also exist to add additional
low level alarms on the plant process computer. In addition to these
alarms, control room operators have twelve other indications of
insufficient, or no, AFW flow to the steam generators. These
indications include three auxiliary feedwater pump low discharge
pressure alarms (one per AFW pump), two auxiliary feedwater flow
meters (one per steam generator), two AFW pump motor amp meters (one
per motor-driven AFW pump), two ``ESF in Pullout'' alarms (one per
Engineered Safety Features train) and three pump running lights (one
per AFW pump). The ten minutes for operator action was discussed in
a telephone conversation between WPSC and Mr. R. Laufer (NRR). Ten
minutes for operator action is further supported by Branch Technical
Position EISCB 18. Scenarios have been completed on the KNPP
simulator to support ten minutes for operator initiation of AFW
flow. In all cases, operators manually initiated AFW flow within the
allowed ten minutes.
Proposed TS 3.4.b.6.C permits valves AFW-10A and AFW-10B to be
in the closed position when below 15% reactor power without
declaring the turbine-driven AFW train inoperable. This change is
being proposed to allow operational flexibility of the AFW system
during startups and shutdowns. As described below, the operability
of the turbine-driven auxiliary feedwater train is independent of
the position of the valves AFW-10A and AFW-10B. However, the
operability of this train is dependent on the ability of these
valves to reposition.
The operability of the AFW system following a main steam line
break (MSLB) was reviewed in our response to IE Bulletin 80-04. As a
result of this review, requirements for the turbine-driven AFW pump
were originally added to the Technical Specifications.
For all other design basis accidents, the two motor-driven AFW
pumps supply sufficient redundancy to meet single failure criteria.
In a secondary line break, it is assumed that the pump discharging
to the intact steam generator fails and that the flow from the
redundant motor-driven AFW pump is discharging out the break.
Therefore, to meet single failure criteria the turbine-driven AFW
pump was added to Technical Specifications.
The cross-connect valves (AFW-10A and AFW-10B) are normally
maintained in the open position. This provides an added degree of
redundancy above what is required for all accidents except for a
MSLB. During a MSLB, one of the cross-connect valves will have to be
repositioned regardless if the valves are normally open or closed.
Therefore, the position of the cross-connect valves does not affect
the operability of the turbine-driven AFW train. However,
operability of the train is dependent on the ability of the valves
to reposition.
For these reasons, this change will have no adverse effect on
the health and safety of the public or significantly increase the
probability or consequences of an accident previously evaluated in
the USAR.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The auxiliary feedwater system is required to mitigate the
consequences of an accident. The auxiliary feedwater system is not
an accident initiator. Therefore, the proposed change does not
create the possibility of a new or different kind of accident.
3. Involve a significant reduction in the margin of safety.
This change alters the assumptions of the safety analysis for
the Small-Break Loss-of-Coolant Accident, the Steam Generator Tube
Rupture and the Loss of Normal Feedwater due to their dependence on
the AFW system to start and supply AFW flow for heat removal. To
support this change the Westinghouse Electric Corporation has
performed an analysis of the Small-Break Loss-of-Coolant Accident
using the NOTRUMP code assuming ten minutes for operator action to
initiate auxiliary feedwater. This analysis resulted in a Peak
Cladding Temperature (PCT) of 1053 deg. F from an initial power
level of 100%. In addition, all other acceptance criteria of 10 CFR
50.46 were met. This large margin to the 2200 deg. F PCT limit
supports ten minutes for operator action to initiate auxiliary
feedwater.
Furthermore, WPSC has analyzed the Loss of Normal Feedwater and
the Steam Generator Tube Rupture Accident assuming delays in the
initiation of auxiliary feedwater. The Loss of Normal Feedwater
Accident with a ten-minute delay in the initiation of Auxiliary
Feedwater does not result in any adverse condition in the core.
[[Page 58409]]
It does not result in water relief from the pressurizer safety valves,
nor does it result in uncovering the tube sheets of the steam
generators. Also, at all times the Departure from Nucleate Boiling
Ratio (DNBR) remained greater than 1.30. The Steam Generator Tube
Rupture Accident with no Auxiliary Feedwater flow was also analyzed.
The results of this analysis indicate that neither steam generator
empties of liquid and at least 20 deg. F of reactor coolant system
subcooling is maintained throughout the transient. Also, there is no
increase in the radiological dose to the public. For these reasons,
these changes will not adversely affect the health and safety of the
public or involve a significant reduction in the margin of safety.
As discussed in the safety evaluation, the operability of the
turbine-driven AFW train is independent of the position of valves
AFW-10A and AFW-10B. However, the operability of the train is
dependent on the ability of these valves to be repositioned.
Therefore, the proposed change has no impact on the accident
analysis and no effect on the margin of safety.
Significant Hazards Determination for Proposed Administrative
Changes to Section TS 3.4, ``Steam and Power Conversion System''
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will 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 the margin of safety.
The proposed changes are administrative in nature and do not
alter the intent or interpretation of the TS. Therefore, no
significant hazards exist.
Additionally, the proposed change is similar to example C.2.e(i)
in 51 FR 7751. Example C.2.e.(i) states that changes which are
purely administrative in nature; i.e., to achieve consistency
throughout the Technical Specifications, correct an error, or a
change in nomenclature, are not likely to involve a significant
hazard.
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 Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
PO Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: Gail H. Marcus.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 18, 1995.
Description of amendment request: This license amendment would
replace the current fuel oil volume requirement in the emergency diesel
generator (EDG) day tank in Technical Specifications 3.8.1.1.b.1) and
3.8.1.2.b.1) with a fuel oil level requirement. Associated Surveillance
Requirement 4.8.1.1.2.a.1) would also be changed to replace the
requirement to visually check the fuel oil level in the day tank with a
requirement to verify that the fuel oil transfer pump starts on low
level in the day tank standpipe.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change will increase the minimum amount of diesel
fuel oil that the current specifications require to be maintained in
the EDG day tanks for standby operation. This change reflects the
level that has been administratively maintained since the beginning
of plant operation. The proposed change will not affect the way the
EDG is operated and does not affect the ability of the EDGs to
perform their safety function. The surveillance requirement change
is being made to more thoroughly reflect the method used to assure
the tank level is being properly maintained. The proposed change
will not require the EDG to be operated in a manner different than
that for which it was designed. Therefore, the proposed change will
not significantly increase the consequences of an accident or
malfunction of equipment important to safety previously evaluated in
the USAR.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no active components being added whose failure could
prevent the EDG from functioning. There is no new type of accident
or malfunction being created and the method and manner of plant
operation remains unchanged. The safety design bases in the USAR
have not been altered. Thus, this change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
No new or different accident scenarios, transient precursors,
failure mechanisms, or limiting single failures will be introduced
as a result of these changes. The method of operation of the EDGs is
not being altered, and the fuel oil transfer pumps will continue to
perform the same function they currently perform. Therefore, the
possibility of a new or different kind of accident other than those
already evaluated will not be created by this change.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There are no changes being made to any safety limits or safety
system settings that would adversely impact plant safety. Although
the minimum required amount of fuel oil specified in the Technical
Specifications is being revised, this amount of fuel oil has been
administratively controlled since the beginning of commercial
operation. Thus, the operability of the emergency diesel generators
has never been affected by this issue. Neither the method of
operation of the EDGs nor their safety function are being altered by
the proposed change. Therefore, the proposed change would not result
in a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 24, 1995.
Description of amendment request: This license amendment request
proposes to revise Surveillance Requirement 4.7.6.e.4 to reflect a
design change, scheduled to be installed during the next refueling
outage, that would change the output rating of the charcoal filter
adsorber unit heater in the pressurization portion of the control room
emergency ventilation system (CREVS) from 15 kW to 5 kW. Proposed
revisions to Surveillance Requirements 4.7.6.c.2 and 4.7.6.d are
included which would change the acceptance criteria for the testing of
carbon samples from the CREVS charcoal adsorbers. The proposal would
adapt ASTM D 3803-1989 as the laboratory testing standard with the
testing to be performed at 30 degrees Centigrade and 70 percent
[[Page 58410]]
relative humidity for a methyl iodide penetration of 2 percent.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The design function of the filter adsorber unit heater in the
pressurization system portion of CREVS is to reduce the relative
humidity of the air entering the charcoal filter beds to 70%
relative humidity. Although the original design specified a heater
with a rating of 15 kW, review of the design basis calculation for
this system indicates that only 2.09 kW is actually required
(including applicable margins to allow for voltage variations). The
proposed change to the CREVS heaters' output rating from 15 kW to 5
kW will not affect the method of operation of the system, and the
new heater capacity will still exceed filter operational
requirements and safety margin. Neither the heater change nor the
charcoal testing protocol changes will affect system operation or
performance, nor do they affect the probability of any event
initiators. These changes do not affect any Engineered Safety
Features actuation setpoints or accident mitigation capabilities.
Therefore, the proposed changes will not significantly increase the
consequences of an accident or malfunction of equipment important to
safety previously evaluated in the USAR.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The requested change to the CREVS heaters' output rating and the
changes to the charcoal sample testing protocol will not affect the
method of operation of the system, and the new heater capacity will
still exceed filter operational requirements and safety margin by a
significant amount. The proposed changes only affect the heater size
in the system and the testing criteria for the charcoal samples. No
new or different accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of these changes. Therefore, the possibility of a new or
different kind of accident other than those already evaluated will
not be created by this change.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The requested change to the CREVS heaters' output rating will
reduce the heater output of the system, but the new heater capacity
will still exceed filter operational requirements and safety margin
by a significant amount. In addition, the reduction in heat load
output from the heater will increase the design margin between the
cooling capacity of the system air conditioning units and the
building heat load. The new charcoal adsorber sample laboratory
testing protocol is more stringent than the current testing practice
and more accurately demonstrates the required performance of the
adsorbers following a design basis LOCA [loss-of-coolant accident].
Therefore, these changes will not reduce the margin of safety of the
CREVS filter operation.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
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.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: September 20, 1995.
Description of amendment request: The proposed amendment would
modify the Appendix A Technical Specifications for the Engineered
Safety Features Actuation System (ESFAS) Instrumentation. Specifically,
the proposed amendment would revise the Seabrook Station Technical
Specifications to relocate Functional Unit 6.b, ``Feedwater Isolation--
Low RCS Tavg Coincident with a Reactor Trip'' from Technical
Specification 3.3.2. ``Engineered Safety Features Actuation System
Instrumentation'' to the Seabrook Station Technical Requirements Manual
which is a licensee controlled document.
Date of publication of individual notice in Federal Register:
October 24, 1995 (60 FR 54524).
Expiration date of individual notice: November 24, 1995.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
[[Page 58411]]
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: October 25, 1994, as
supplemented by letter dated September 11, 1995.
Brief Description of amendments: The proposed amendments change the
Technical Specifications to relocate the remaining Environmental
Technical Specifications to other licensee-controlled documents and
delete the 30-day reporting requirement for inoperable meteorological
instrumentation.
Date of issuance: November 2, 1995.
Effective date: November 2, 1995.
Amendment Nos.: 179 and 210.
Facility Operating License Nos. DPR-71 and DPR-62. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 1994 (59 FR
63113). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: May 13, 1993 as supplemented
August 11 and September 20, 1995.
Brief description of amendments: The amendments revised Section 3/
4.6.1.7 of the Technical Specifications, Containment Purge Ventilation
System, to allow the simultaneous opening of the 8-inch miniflow purge
supply and exhaust valves to ensure the containment atmosphere is
conducive to human occupants and to maintain their dose as low as
reasonably achievable.
Date of issuance: November 2, 1995.
Effective date: November 2, 1995.
Amendment Nos.: 76, 76, 68, and 68.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 15, 1993 (58
FR 48379). The August 11 and September 20, 1995, submittals provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 2, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: September 1, 1995, as
supplemented on September 1 (two letters), September 2, September 4,
September 8, September 15, September 19, September 20, September 22,
October 3, October 7, October 11 (two letters), October 13 (three
letters), October 23 and October 26, 1995.
Brief description of amendments: The amendments revise the steam
generator (SG) repair criteria in the Byron, Unit 1 and Braidwood, Unit
1 Technical Specifications. These revisions add a set of voltage-based
SG tube repair criteria different from those previously added by
License Amendment No. 66, dated October 24, 1994, to the Byron 1 TSs
and by License Amendment No. 54, dated August 18, 1994, to the
Braidwood 1 TSs. The present set of voltage repair limits which are
being added to the Byron 1 and Braidwood 1 TSs are applicable only for
a specific form of SG tube degradation identified as outer diameter
stress corrosion cracking (ODSCC) which is confined entirely within the
thickness of the tube support plates (TSPs) in the SGs. The voltage-
based repair criteria for the cold-leg side of the SGs for SG tubes
with ODSCC indications and for SG tubes on the hot-leg side which show
significant denting, are consistent with those provided in the NRC
staff's guidance contained in Generic Letter 95-05, dated August 3,
1994.
The lower voltage repair limit for the SG tubes with ODSCC
indications on the hot-leg side of the SGs have been raised from 1.0 to
3.0 volts as measured by a bobbin coil. All bobbin indications below
3.0 volts will be allowed to remain in service and all bobbin
indications above this limit will be either repaired or removed from
service by plugging.
This revision to the voltage repair limits on the hot-leg side
reflects a methodology which is significantly different than that
contained in GL 95-05. The principal difference between the methodology
being applied for the 3.0 volt criteria on the hot-leg side is that the
Commonwealth Edison Company (ComEd) is taking credit for the constraint
provided by the TSPs to reduce the probability of SG tube burst in the
event of a severe accident (i.e., a main steamline break). This
constraint is assured by modifying a limited number of SG tubes so that
they provide additional stiffness to the TSPs, thereby reducing to a
small amount, their deflection under MSLB blowdown loads.
Additionally, inspection and reporting requirements are being added
to the Byron 1 and Braidwood 1 TSs in support of the revised voltage-
based repair criteria. Further, the maximum permissible value of the
iodine-131 concentration in the primary coolant in the Byron 1 TSs is
reduced from 1.0 to 0.35 microcuries per gram of coolant. This is the
same value for the iodine-131 primary coolant concentration in the
Braidwood 1 TSs. Finally, the Bases sections in the Byron 1 and
Braidwood 1 TSs are revised to provide a concise description of the
methodology proposed by ComEd in support of its proposed revision of
the voltage-based SG tube repair criteria.
Date of issuance: November 9, 1995.
Effective date: November 9, 1995.
Amendment Nos.: 77, 77, 69, and 69.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49963).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 9, 1995. The supplemental
submittals listed above provide clarifying technical information that
does not affect the initial No Significant Hazards Consideration
Determination.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
[[Page 58412]]
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: July 5, 1995.
Brief description of amendment: This amendment revises Section 6.0
of the Technical Specifications to incorporate several administrative
controls and editorial changes to the Training, Plant Review Committee,
and Plant Safety and Licensing staff sections.
Date of issuance: November 3, 1995.
Effective date: November 3, 1995.
Amendment No.: 170.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39435).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: April 10, 1995.
Brief description of amendments: The amendments revise the required
number of operable hydrogen igniters to allow removal of two hydrogen
igniters serving the lower reactor cavity and incore instrument cable
tunnel.
Date of issuance: October 30, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 136 and 130.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49932).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 30, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 13, 1995.
Brief description of amendments: The amendments modify the notation
for the overpower delta temperature reactor trip heatup setpoint
penalty coefficient as delineated in Note 3 in Technical Specification
Table 2.2-1 in order to make the nomenclature consistent with the
Standard Technical Specifications and to facilitate a modification to
reduce the reactor coolant system hot leg temperature as planned during
the Catawba Unit 2 end-of-cycle 7 refueling outage.
Date of issuance: October 31, 1995.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: 137 and 131.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49933).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 1, 1995, as
supplemented October 17, 1995.
Brief description of amendments: The amendments revise Technical
Specification (TS) 6.9.1.9 to include references to updated or recently
approved methodologies used to calculate cycle-specific limits
contained in the Core Operating Limits Report. The subject references
have been reviewed and approved by the NRC staff.
Date of issuance: November 2, 1995.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: 138 and 132.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49932). The October 17, 1995, letter provided clarifying information
that did not change the scope of the September 1, 1995 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 November 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
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: June 13, 1994, as supplemented
by letters dated August 15, 1994, March 23, April 18, July 21, and
September 22, 1995.
Brief description of amendments: The amendments revise the
Technical Specifications to increase the initial fuel enrichment limit
and establish new loading patterns for new and irradiated fuel in the
spent fuel pool to accommodate this increase.
The March 23, 1995, supplement, which provided additional
information that modified the June 13, 1994, application's no
significant hazards consideration determination, also revises the TS to
(1) change the surveillance requirement for boron concentration in the
spent fuel pool (SFP), (2) remove the option to use alternate storage
configurations in the SFP and replace it with footnotes, (3) add
information contained in the Bases to the footnotes, and (4) change the
Bases to discuss the option to use specific analyses on alternate fuel.
Date of issuance: November 6, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 159 and 141.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8746); and May 8, 1995 (60 FR 22590). The April 18, July 21, and
September 22, 1995, letters provided additional clarifying information
that did not change the scope of the June 13, 1994, 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 November 6, 1995, and Environmental
Assessment dated August 17, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
[[Page 58413]]
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of application for amendment: May 17, 1995.
Brief description of amendment: The amendment will extend the
applicability of the current Reactor Coolant System (RCS) Pressure/
Temperature Limits and maximum allowed RCS heatup and cooldown rates to
23.6 Effective Full Power Years (EFPY) of operation. In addition,
administrative changes were proposed for TS 3.1.2.1 (Boration Systems
Flow Paths-Shutdown) and TS 3.1.2.3 (Charging Pump-Shutdown) to clarify
the conditions for which a High Pressure Safety Injection pump may be
used.
Date of Issuance: October 27, 1995.
Effective Date: October 27, 1995.
Amendment No.: 141.
Facility Operating License No. DPR-67: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32362).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: February 28, 1994.
Brief description of amendments: The amendments delete the minimum
frequency criteria prescribed for quality assurance audits from
Administrative Controls sections 6.5.2.8 and 6.8.4 of the Technical
Specifications (TS). Audit periodicity will thereby be controlled by
the program described in the Florida Power and Light Company (FPL)
Topical Quality Assurance Report.
Date of Issuance: October 25, 1995.
Effective Date: October 25, 1995.
Amendment Nos.: 140 and 80.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17599).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 25, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida
Date of application for amendments: July 26, 1995.
Brief description of amendments: These amendments revise selected
line items from NRC Generic Letter 93-05, ``Line-Item Technical
Specification Improvements to Reduce Surveillance Requirements for
Testing During Power Operation.''
Date of issuance: October 17, 1995.
Effective date: October 17, 1995.
Amendment Nos.: 177 and 171.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47617).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 17, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 10, 1993.
Brief description of amendment: The amendment revises the Cooper
Nuclear Station Technical Specifications to change the reporting
frequency of the Radioactive Materials Release Report from semiannual
to annual and to extend the reporting frequency of the Annual Design
Change Report from annual to annually or along with the Updated Safety
Analysis Report updates required by 10 CFR 50.71(e). This change
reflects revised requirements contained in 10 CFR 50.36a and 10 CFR
50.59(b).
Date of issuance: November 3, 1995.
Effective date: November 3, 1995.
Amendment No.: 172.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7691).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated Novemver 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 28, 1995.
Brief description of amendment: The amendment revises the Cooper
Nuclear Station Technical Specifications to increase the required
reactor pressure vessel boron concentration, to modify the surveillance
frequency for standby liquid control system pump operability testing
from monthly to quarterly, and to make editorial changes.
Date of issuance: November 8, 1995.
Effective date: November 8, 1995.
Amendment No.: 173.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39441).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 8, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: September 5, 1995.
Description of amendment request: The amendment modifies the
Appendix A Technical Specifications (TSs) for the Turbine Cycle Safety
Valves. Specifically, the amendment changes Seabrook Station Appendix A
Technical Specification Table 3.7-1 to reduce the Maximum Allowable
Power Range Neutron Flux--High Setpoints with Inoperable Main Steam
Safety Valves (MSSVs) and Table 3.7-2 to reduce the opening setpoints
of the MSSVs. Bases Section 3/4.7.1.1 is changed to include the
algorithm used for determining the new setpoint values.
Date of issuance: November 2, 1995.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 43.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 2, 1995 (60 FR
51505).
[[Page 58414]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, New Hampshire 03833.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: December 21, 1994, as
supplemented February 22, 1995.
Brief description of amendment: The amendment revises the License
Condition C.(3), Fire Protection, and certain of the Technical
Specifications (TS) related to fire protection requirements. The
amendment changes the TS by relocating them to another controlled
document, the Technical Requirements Manual referenced in the Final
Safety Analysis Report.
Date of issuance: November 3, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 191.
Facility Operating License No. DPR-65: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6303) The February 22, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: August 31, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to remove the phrase ``other than Millstone Unit No. 2''
from the Administrative Controls Section 6.3.1, Item (a). This relates
to Amendment No. 178 that changed the Technical Specifications to
require an individual who serves as the Operations Manager to either
hold a Millstone Unit 2 Senior Reactor Operator (SRO) license or have
held an SRO license at another pressurized water reactor other than the
Millstone Unit No. 2. If the Operations Manager does not hold a
Millstone Unit No. 2 SRO license, then an individual serving as the
Assistant Operations Manager would be required to possess an SRO
license at Millstone Unit 2.
Date of issuance: November 2, 1995.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 190.
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49941).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: January 27, 1995.
Brief description of amendments: The amendments change the Limerick
Generating Station Units 1 and 2 Technical Specifications (TS) by
eliminating the TS active safety function designation of eight (i.e.,
four per unit) Drywell Chilled Water System valves.
Date of issuance: October 30, 1995.
Effective date: October 30, 1995.
Amendment Nos.: 103 and 67.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20524).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 30, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: November 23, 1994, as
supplemented by letter dated August 31, 1995.
Brief description of amendment: The proposed changes to the
Technical Specifications (TSs) revise TS 4.8.2.1, ``Electrical Power
Systems--D.C. Sources,'' Surveillance Requirements, and associated
Bases Section 3/4.8.2.
Date of issuance: October 31, 1995.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 87.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39449). The August 31, 1995, letter provided additional and clarifying
information that did not change the scope of the November 23, 1994,
application and the initial proposed no significant consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: November 28, 1994.
Brief description of amendment: This amendment revises the
technical specifications for the Reactor Coolant System recirculation
flow upscale trip function to change the trip setpoint and allowable
value to reflect 105% of rated core flow, item one of the above
application.
Date of issuance: October 31, 1995.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 86.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
[[Page 58415]]
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39450).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
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: March 30, 1995, as supplemented
August 18, 1995.
Brief description of amendments: The amendments eliminate the
defined term CONTROLLED LEAKAGE, remove Controlled Leakage flow from
the Reactor Coolant System Operational Leakage Limiting Condition for
Operation (LCO) and establish a new Seal Injection Flow LCO.
Date of issuance: October 30, 1995.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 178 and 159.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24918). The August 18, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 30, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: August 1, 1995, as supplemented
by letter dated October 18, 1995.
Brief description of amendments: These amendments revise Technical
Specification (TS) 3/4.3.2, ``Engineered Safety Features Actuation
System Instrumentation,'' Table 3.3-3. Table 3.3-3 includes the
requirements for the minimum number of toxic gas isolation system
(TGIS) trains operable. These amendments are a one-time-only change to
extend the allowed TGIS outage times during the replacement of the
existing TGIS instrumentation.
Date of issuance: November 2, 1995.
Effective date: November 2, 1995, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--Amendment No. 126; Unit 3--Amendment No.
115.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47625). The October 18, 1995, supplemental letter provided
clarifying information and did not change the initial no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
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: September 30, 1993 (TS-337).
Brief Description of amendment: The amendments revise the operating
license to reflect issuance of a safety evaluation dated November 2,
1995 accepting the revised Appendix R Safe Shutdown Program to
accommodate simultaneous power operation of Browns Ferry Units 2 and 3.
Date of issuance: November 2, 1995.
Effective Date: November 2, 1995.
Amendment Nos.: 226, 241 and 200.
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 5, 1994 (59 FR
629).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 1995.
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. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: January 4, 1995 (TS 355).
Brief Description of amendment: The amendments revise applicability
and surveillance requirements for the intermediate power range monitor,
average power range monitor (APRM), and APRM Inoperative Trip
functions.
Date of issuance: November 2, 1995.
Effective Date: November 2, 1995.
Amendment Nos.: 227, 242 and 201.
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29888).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 1995.
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. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: June 2, 1995 (TS 361/371).
Brief Description of amendment: The amendments revise the
operability definition for residual heat removal service water
components for use as a standby coolant supply. The amendments also
incorporate related changes to the technical specification Bases which
were submitted on October 2, 1995.
Date of issuance: November 2, 1995.
Effective Date: November 2, 1995.
Amendment Nos.: 225, 240 and 199.
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42610).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room Location: Athens Public library, South
Street, Athens, Alabama 35611.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of application for amendment: May 19, 1995; revised September
11, 1995 (TS 95-13).
[[Page 58416]]
Brief description of amendment: The amendment modifies License
Condition 2.C.(17) by extending the required surveillance interval to
May 18, 1996, for Surveillance Requirement 4.3.2.1.3 for certain
specified engineered safety features response time tests.
Date of issuance: October 30, 1995.
Effective date: October 30, 1995.
Amendment No.: 204.
Facility Operating License No. DPR-79: Amendment revises the
operating license.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32372); renoticed September 27, 1995 (60 FR 49948).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Dated at Rockville, Maryland, this 15th day of November 1995.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 95-28606 Filed 11-24-95; 8:45 am]
BILLING CODE 7590-01-P