[Federal Register Volume 59, Number 81 (Thursday, April 28, 1994)]
[Unknown Section]
[Page ]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-10011]
[Federal Register: April 28, 1994]
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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 April 2, 1994, through April 15, 1994. The
last biweekly notice was published on April 13, 1994 (59 FR 17591).
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 P-223, Phillips Building, 7920 Norfolk Avenue,
Bethesda, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies
of written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street NW., Washington, DC 20555. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By May 27, 1994, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors:
(1) The nature of the petitioner's right under the Act to be made a
party to the proceeding;
(2) The nature and extent of the petitioner's property, financial,
or other interest in the proceeding; and
(3) The possible effect of any order which may be entered in the
proceeding on the petitioner's interest. The petition should also
identify the specific aspect(s) of the subject matter of the proceeding
as to which petitioner wishes to intervene. Any person who has filed a
petition for leave to intervene or who has been admitted as a party may
amend the petition without requesting leave of the Board up to 15 days
prior to the first prehearing conference scheduled in the proceeding,
but such an amended petition must satisfy the specificity requirements
described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): Petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street NW., Washington, DC 20555, and at the local public document room
for the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of Amendment Requests: January 4, 1994.
Description of Amendment Requests: The proposed amendment would
change Technical Specification 3/4.2.3 Azimuthal Power Tilt and its
associated bases. The licensee proposed to change the Azimuthal Power
Tilt limit from less than or equal to 10 percent to less than or equal
to 3 percent when the Core Operating Limit Supervisory System is out of
service.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Decreasing the COLSS [Core Operating Limit Supervisory System]
out-of service Azimuthal Power Tilt Technical Specification limit
does not increase the probability or consequences of an accident
previously evaluated. The Technical Specification operating limit is
being conservatively reduced to conform to the assumptions used in
the safety analysis. The reduced operating limit requires a more
uniform power distribution in the reactor core. The uniform power
distribution may reduce the consequences of an accident previously
evaluated by not allowing regions in the core to operate at higher
power levels.
Standard 2--Create the possibility of a new or different kind of
accident from any accident previously analyzed.
The proposed amendment will result in an alarm setpoint change,
but does not involve any equipment changes and will not alter the
manner in which the plant will be operated. For this reason, this
amendment will not create the possibility of an new or different
kind of accident from any previously evaluated. The proposed
operating range is smaller and completely within the existing
Technical Specification limits; thus, there are no mechanisms to
create the possibility of a new or different kind of accident from
those previously evaluated.
Standard 3--Involve a significant reduction in a margin of
safety.
The proposed amendment conservatively reduces the COLSS out-of-
service Azimuthal Power Tilt Technical Specification limit, thereby
increasing the margin of safety. The proposed operating range is
smaller and completely bounded by the existing Technical
Specification limits.
The NRC staff has reviewed the licensees' analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room Location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Attorney for Licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: Theodore R. Quay.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of Amendment Request: March 25, 1994.
Description of Amendment Request: The amendment would revise
Technical Specification 3/4.8.4.2, Motor Operated Valves Thermal
Overload Protection, with a more accurate description of the motor-
operated valve (MOV) bypass configuration.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This change is administrative in nature, providing a more
accurate description of the MOV electrical supply configuration
related to the thermal overload bypass function. Therefore, the
change in terminology would not increase the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve any modifications or
additions to plant equipment and the design and operation of the
plant will not be affected. Therefore, the change in MOV thermal
overload bypass function terminology would not increase the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed terminology change does not affect any parameters
which relate to the margin of safety as defined in the Technical
Specifications or in the FSAR [Final Safety Analysis Report].
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for Licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman.
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: March 7, 1994, as supplemented
on March 24, 1994.
Description of Amendment Requests: The proposed amendment would
change Technical Specification 4.6.1.2 by removing the specific
schedules for containment integrated leak rate testing (CILRT) and
specifying that the testing will be done in accordance with Appendix J
to 10 CFR part 50.
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 proposed change will allow flexibility in the scheduling for
Type A tests in the 10-year service period while still meeting the
requirements in 10 CFR 50 Appendix J. Additional flexibility is
needed for plants using an 18-month fuel cycle to allow refueling
outages and 10-year inservice testing intervals to coincide. For
performance of the third Type A test at Byron, the change would
allow an extension of four (4) months beyond the current maximum 50-
month surveillance interval. The third test would be completed at
the fifty-four (54) month interval for Byron Units 1 and 2.
For Braidwood Units 1 and 2, an extension on the surveillance
time interval will not be necessary to satisfy the requirements of
Appendix J. The Braidwood Units have scheduled the third Type A test
to be conducted with the 10-year Inservice Inspection.
The results of the previous Type A leak tests show the overall
leakage from the Byron containment buildings at very low levels. The
extension of the Type A test by four months would not cause the
consequences of a previously evaluated accident to increase. By
continuing to conform to the requirements of 10 CFR 50 Appendix J,
the test frequency, methodology, and acceptance criteria for
containment leakage remains the same. Therefore, there is no
significant increase in the probability or the 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 do not affect the design or operation of
any system, structure or component in the plant. There are no
changes to parameters governing plant operation and no new or
different type of equipment will be installed. No new accident
scenarios are created by the proposed change because the test
frequency continues to meet the requirements of Appendix J of 10 CFR
part 50. There is no affect on containment structure, the
penetrations, or the facility. The proposed change to the test
schedule only provides flexibility in meeting the same requirement
for three tests in a 10-year period. The testing method and bases
have not changed. Therefore, operation of the units with this more
flexible test schedule will not result in an accident previously not
analyzed in the Updated Final Safety Analysis Report (UFSAR). The
proposed changes do not impact the design bases of the containment
and do not modify the response of the containment during a design
basis accident. Therefore, the changes do not create the possibility
of a new or different type of accident from any accident previously
evaluated.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the margin of safety for any
Technical Specifications. The initial conditions and methodologies
used in the accident analyses remain unchanged, therefore, the
results of the accident analyses are not impacted. The proposed
change to the schedule allows for additional flexibility in meeting
the requirement for three tests in a 10-year period. Elimination of
the specified time interval for Type A testing would allow Byron
Units 1 and 2 to extend the surveillance requirement of the third
Type A test by four (4) months. This would exceed the existing
maximum 50 month interval currently specified in Technical
Specifications. The extension will allow performance of the Type A
test to coincide with the seventh refueling outage, 10 year
Inservice Inspection, and continue to meet the requirements of
Appendix J to 10 CFR part 50. These proposed changes do not affect
or change any limiting conditions for operation (LCO), or any other
surveillance requirements in the Technical Specifications.
The results of the previous Type A leak tests have shown that
the overall leakage rates from the Byron containment buildings were
at low levels. The latest test results for Units 1 and 2 were 0.0175
weight percent per day and 0.0376 weight percent per day,
respectively. The overall containment leakage rates have
consistently remained well below the acceptance criteria for Byron
Station Type A tests of 0.075 weight percent per day. The testing
method, acceptance criteria, and bases for the surveillance
requirement will not be changed by the proposed amendment.
The present test performance margins, coupled with the Type B &
C test program for monitoring and repairing individual leakage
components provides justification for the proposed change. The Type
B & C tests provide added assurance that the overall containment
integrated leakage rates remain satisfactory. No significant leakage
trends have been identified which threaten the overall containment
leakage specifications.
In summary, Commonwealth Edison concludes that this change does
not involve a significant reduction in a margin of safety because
the containment integritiy will be maintained. Testing in accordance
with Appendix J requirements ensures confidence is containment
intergity. The proposed Technical Specifications amendment will
continue to require testing that is consistent with Appendix J
requirements. Additionally, results from previous tests have shown
acceptable low overall containment leakage rates. Extension of Type
A testing for four months would not involve a signficant 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: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township 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 60690.
NRC Project Director: James E. Dyer.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York Date
of Amendment Request: February 18, 1994
Description of Amendment Request: This amendment is an additional
followup to the amendment request of May 29, 1992, published in the
Federal Register on July 8, 1992 (57 FR 30242), which changed the
Technical Specifications Sec. 1.0, Definitions, to accommodate a 24-
month fuel cycle and which proposed the extension of the test intervals
for specific surveillance tests. This amendment proposes extending the
surveillance intervals to 24 months for the following additional
surveillance tests:
(1) Analog Rod Position Indication.
(2) Plant Noble Gas Activity Monitor (R-44).
(3) Low Turbine Auto Stop Oil Pressure Reactor Trip.
(4) 6.9 KV Undervoltage Relays.
(5) Boric Acid Tank Level.
(6) Vapor Containment Sump Discharge Flow and Temperature
Channel.
(7) Loss of Power Undervoltage and Degraded Voltage Relays.
(8) Over-pressurization Protection System (OPS) and Control Rod
Protection System (for use with Low Parasite [LOPAR] fuel) Trip.
(9) Condenser Evacuation System Activity Monitor (R-45).
(10) Service Water Inlet Temperature Monitoring Instrumentation.
(11) Sampler Flow Rate Monitors.
(12) Boric Acid Makeup Flow System.
(13) Plant Vent Noble Gas Effluent Monitor (R-27).
The amendment also proposes to change the surveillance interval for
the Refueling Water Storage Tank Level to quarterly and to change the
trip setpoint for the Control Rod Protection System. The changes
requested by the licensee are in accordance with Generic Letter 91-04,
``Changes in Technical Specification Surveillance Intervals to
Accommodate a 24-Month Fuel Cycle.''
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[(1) Analog Rod Position Indication:]
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
analog rod position indication channel be changed from every 18
months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that none of the major error contributors are time
dependent and that it can be reasonably expected that the channel
will remain within calibration tolerance over a possible 30 month
operating cycle. In addition, the rod bottom bistable is subject to
monthly testing which would detect any abnormalities in an extended
operating cycle. Due to this monthly test and the acceptable past
test history, it is concluded that the channel will continue to
operate within tolerance over an extended operating cycle and will
not contribute to a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval is not expected to affect the
ability of the instrument channel to remain within calibration
tolerance. Furthermore, the rod position indicator is used in normal
operation only as an aid in control rod movement. Normally, very
little control rod movement occurs during normal operation.
Furthermore, it is not relied upon for accident prevention or
accident mitigation. In accordance with existing Technical
Specifications, normal operation can continue even if one channel is
inoperable because alternate means (core instrumentation) exists to
monitor rod position. The frequent monthly test tends to minimize
the effect of a longer operating cycle for the rod position
indication channel as any malfunction induced by time would be
detected. Thus, it is concluded that the possibility of a new or
different kind of accident from any accident previously evaluated
has not been created.
3. A significant reduction in a margin of safety is not
involved.
A statistical analysis of past calibration data has not
identified any time dependent error contributors. Also, past test
data indicates that the channel remains within calibration tolerance
over the existing operating cycle. A longer operating cycle would
increase the risk of drift, however accuracy is not a prime
requirement for the RPI. Therefore, it is concluded that a longer
operating cycle will not result in a significant reduction in a
margin of safety.
[(2) Plant Noble Gas Activity Monitor (R-44):]
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Plant Noble Gas Activity Monitor (R-44) be changed from every 18
months (+25%) to every 24 months (+25%).
The function of R-44 is to respond to high activity levels
during normal operation.
The setpoint for R-44 is established sufficiently above the
expected radioactivity level in the discharge stream to preclude
false actions but sufficiently below the allowed discharge
radioactivity concentration so that discharge in excess of
permissible limits does not occur. Monitor readouts are not used for
quantitative purposes, but are used to respond to relative changes
in radioactivity concentration.
There is limited data to support an unqualified extension of the
surveillance interval. However, the instrument is checked for
operability prior to release. Should the instrument be inoperable
releases may continue provided grab sample analysis is performed.
Since the monitor is subject to daily channel checks, monthly source
checks, and quarterly functional channel tests, abnormal instrument
behavior or inoperability would be detected permitting corrective
actions during the extended surveillance interval.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Operability of the instrument is important rather than ability
to maintain a specific setpoint. Operability of the instrument is
verified prior to a planned discharge and this is independent of an
extended surveillance cycle.
3. There has been no reduction in the margin of safety.
As the Technical Specifications permit pre-planned release even
with an inoperable instrument, the margin of safety is not impacted
by an extended surveillance interval provided that instrument
operability is verified prior to release. This is also required by
the Technical Specifications.
[(3) Low Turbine Auto Stop Oil Pressure Reactor Trip:]
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Low Turbine Auto Stop Oil Pressure system be changed from every 18
months (+25%) to every 24 months (+25%).
No credit is taken for a reactor trip from a low turbine auto
stop oil pressure signal resulting from a turbine trip. Rather, the
safety analysis assumes this reactor trip does not occur during full
load rejection until an overpower delta T condition causes a reactor
trip. In addition, no credit is taken for this system for turbine
missile protection. Therefore, extending the surveillance interval
for this parameter has no impact upon the probability or
consequences of an accident.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
As no credit is taken in the safety analysis for this trip, the
possibility of a new or different kind of accident has not been
created by extending the surveillance interval.
3. There has been no reduction in the margin of safety.
Past test results have not identified any failures. Therefore,
pursuant to Generic Letter 91-04, it is reasonably expected that
this system will continue to function in an acceptable manner over
an extended operating cycle.
[(4) 6.9 kv Undervoltage Relays:]
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the calibration frequency for the 6.9 kv
undervoltage channel be changed from every 18 months (+25%) to every
24 months (+25%).
Quarterly testing of these relays is required by Technical
Specifications. The data from the quarterly tests of the new relays
will be used to assure that drift does not exceed projected values.
The quarterly tests provide a means of maintaining calibration
within specified values, virtually eliminating any impact upon
safety from an extended operating cycle.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Because the quarterly tests assure that relay performance
remains within specified limits, there is no possibility of creating
a new or different kind of accident from any previously analyzed.
3. There has been no reduction in the margin of safety.
The requirement for a channel functional test each quarter
minimizes any potential impact upon safety due to an extended
operating cycle.
[(5) Boric Acid Tank Level:]
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequence of
an accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
Boric Acid Tank Level instrumentation be changed from every 18
months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists between the existing
Technical Specification limit and the licensing basis Safety
Analysis limit to accommodate the channel statistical error
resulting from a 30 month operating cycle. The existing margin
between the Technical Specification limit and the Safety Analysis
limit provides assurance that plant protective actions will occur as
required. It is therefore concluded that changing the surveillance
interval from 18 months (+25%) to 24 months (+25%) will not result
in a significant increase in the probability or consequences of an
accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds the current margin between the
existing Technical Specification limit and the Safety Analysis
limit. Plant equipment, which will be set at (or more conservatively
than) Technical Specification limits, will provide protective
functions to assure that Safety Analysis limits are not exceeded.
This will prevent the possibility of a new or different kind of
accident from any previously evaluated from occurring.
3. A significant reduction in a margin of safety is not
involved.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the margin which exists between the current
Technical Specification limit and the licensing basis Safety
Analysis limit. This margin, which is equivalent to the existing
margin, is necessary to assure that protective safety functions will
occur so that Safety Analysis limits are not exceeded.
[(6) Vapor Containment Sump Discharge Flow and Temperature
Channel:]
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the calibration frequency for the VC sump
discharge flow and temperature channel be changed from every 18
months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists between the existing
Technical Specification and the licensing basis Safety Analysis to
accommodate the channel statistical error resulting from a 30 month
operating cycle. The existing margin between the Technical
Specification and the Safety Analysis provides assurance that plant
protective actions will occur as required. It is therefore concluded
that changing the surveillance interval from 18 months (+25%) to 24
months (+25%) will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds the current margin between the
existing Technical Specification and the Safety Analysis. Plant
equipment, which will be set at (or more conservatively than)
Technical Specification limits, will provide protective functions to
assure that Safety Analysis limits are not exceeded. This will
prevent the possibility of a new or different kind of accident from
any previously evaluated from occurring.
3. A significant reduction in a margin of safety is not
involved.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the margin which exists between the current
Technical Specification and the licensing basis Safety Analysis.
This margin, which is equivalent to the existing margin, is
necessary to assure that protective safety functions will occur so
that Safety Analysis limits are not exceeded.
[(7) Loss of Power Undervoltage and Degraded Voltage Relays:]
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
The Technical Specifications specify that the Loss of Power
(undervoltage and degraded voltage) relays be calibrated and tested
at a refueling interval; that the undervoltage alarm be calibrated
at a refueling interval, and that the undervoltage (station
blackout) input to Auxiliary Feedwater be calibrated at refueling
intervals. It is proposed that the surveillance frequency be revised
from 18 months (+25%) to 24 months (+25%).
All of the undervoltage and station blackout relays were found
to be within specification at each of the refueling outage
calibration periods.
Since the old relays have been replaced with relays from a
different manufacturer whose drift characteristics are expected to
be superior, extending the surveillance interval by several months
will not significantly increase the probability or consequences of
an accident.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Past test results provide reasonable assurance that the relays
will perform in an acceptable manner for an extended operating
cycle. With the installation of the new relays, whose performance
will surpass the old relays, it is concluded that the plant will
perform within its design basis for an extended operating cycle.
Therefore, the possibility of a new or different kind of accident
from any previously analyzed has not been created.
3. There has been no significant reduction in the margin of
safety.
Since the new relays will surpass the performance of the old
relays, there is reasonable assurance that a significant reduction
in the margin of safety has not resulted from an extended operating
cycle.
[(8) Over-pressurization Protection System (OPS) and Control Rod
Protection System (for use with Low Parasite (LOPAR) fuel) Trip:]
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
Over-pressurization protection system and the LOPAR trip system be
changed from every 18 months (+25%) to every 24 months (+25%). This
necessitates a change in the LOPAR Technical Specification trip
setpoint from 350 deg.F to 381 deg.F.
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed based upon historical test data.
Based on this analysis, a change to the Technical Specifications is
required. Sufficient margin exists between the Safety Analysis limit
and the proposed Technical Specification limit to accommodate
projected channel uncertainty over a 30 month operating cycle. A
statistical basis exists to assure that protective action will occur
to prevent Safety Analysis limits from being exceeded. Thus, there
will not be a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident
previously evaluated has not been created.
Based upon a statistical analysis of past historical test data
it has been demonstrated that reasonable assurance exists to
conclude that Safety Analysis limits will not be exceeded over a 30
month operating cycle. The proposed Technical Specification limits
provide margin with respect to the Safety Analysis limits and
confidence that appropriate plant protective response will be
provided to prevent the possibility of a new or different kind of
accident from that previously evaluated from being created.
3. A significant reduction in a margin of safety is not
involved.
The proposed changes to the Technical Specification limits are
being made to assure that the previously established margin remains
the same between plant protective function set points and Safety
Analysis limits. This margin is based upon an evaluation of past
historical test data and analytical methods for projecting
instrument channel uncertainty over a 30 month operating cycle. It
is therefore concluded that the existing margin of safety has been
preserved.
[(9) Condenser Evacuation System Activity Monitor (R-45):]
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Condenser Evacuation System Noble Gas Activity Monitor (R-45) be
changed from every 18 months (+25%) to every 24 months (+25%).
Since this radiation monitor is relatively new a degree of
uncertainty is introduced by extending the surveillance interval by
several months. However, the setpoint for automatic diversion is set
some what conservatively. It is established sufficiently high to
avoid spurious actuations and yet sufficiently low so that diversion
and alarm can occur should a step increase in radioactivity level
occur. Under these circumstances considerable departure from the
setpoint can be accommodated and the monitor will still perform its
intended safety function. Continued monitor operability is important
and malfunction would be detected by monthly checks during the
extended operating cycle. Thus, despite the introduction of a new
monitor, the capability of R-45 to tolerate drift in addition to
monthly operator checks, leads to the conclusion that an extended
operating cycle will not result in a significant increase in the
probability or consequences of an accident.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Monthly checks would identify abnormal operating
characteristics, should the instrument fail to perform its intended
function. In the event of tube rupture with a reactor coolant system
radioactivity concentration corresponding to 1% defective fuel, the
resultant site boundary dose would be within 10 CFR [part] 20 limits
should the monitor fail to perform its function (as discussed in
FSAR [Final Safety Analysis Report]). In addition, alternate means
of alarms to indicate a tube rupture event are available. Thus, the
possibility of a new or different kind of accident has not been
created.
3. There has been no reduction in the margin of safety.
Although this monitor is not necessary to mitigate releases
below regulatory limits, it does provide the earliest of a steam
generator tube leak. In this regard, continued instrument operation
is important. Continued instrument operability would be verified by
the monthly checks in an extended operating cycle.
[(10) Service Water Inlet Temperature Monitoring
Instrumentation:]
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
Service Water Inlet Temperature Monitoring Instrumentation be
changed from every 18 months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists between the existing
Technical Specification and the licensing basis Safety Analysis to
accommodate the channel statistical error resulting from a 30 month
operating cycle. The existing margin between the Technical
Specification and the Safety Analysis provides assurance that plant
protective actions will occur as required. It is therefore concluded
that changing the surveillance interval from 18 months (+25%) to 24
months (+25%) will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds the current margin between the
existing Technical Specification and the Safety Analysis. Plant
equipment, which will be set at (or more conservatively than)
Technical Specification limits, will provide protective functions to
assure that Safety Analyses are not exceeded. This will prevent the
possibility of a new or different kind of accident from any
previously evaluated from occurring.
3. A significant reduction in a margin of safety is not
involved.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the existing margin between the current
Technical Specification and the licensing basis Safety Analysis.
This margin, which is equivalent to the existing margin, is
necessary to assure that the protective safety functions occur and
that the Safety Analysis limits are not exceeded.
[(11) Sampler Flow Rate Monitor:]
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Sample Flow Rate Monitors be changed from every 18 months (+25%) to
every 24 months (+25%).
The flow rate monitors are used to estimate the total volume of
air passed through filters. There is no setpoint or safety function
served by these monitors. A high level of radioactivity in the
discharge stream is detected by R-43 and/or R-44.
Insofar as discharge via the unit vent is permissible with the
monitors inoperable, extension of the surveillance interval will
have no impact upon safety.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
As the nuclear safety function is provided by other monitors in
the event of high radioactivity levels in the discharge stream,
extension of the surveillance interval will have no impact upon the
creation of a new or different kind of accident.
3. There has been no reduction in the margin of safety.
These flow monitors are utilized to determine the total air flow
through filters for computational purposes. As adequate measures
(other monitors) exist to prevent the possibility of discharging
radioactivity in excess of applicable limits, there is virtually no
impact upon safety incurred by extending the surveillance interval.
[(12) Boric Acid Makeup Flow System:]
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
Boric Acid Makeup Flow System be revised from every 18 months (+25%)
to every 24 months (+25%). A statistical analysis of channel
uncertainty for a 30 month operating cycle has been performed. Based
upon this analysis it has been concluded that sufficient margin
exists between the existing Technical Specification limit and the
licensing basis Safety Analysis limit to accommodate the channel
statistical error resulting from a 30 month operating cycle. The
existing margin between the Technical Specification limit and the
Safety Analysis limit provides assurance that plant protective
actions will occur as required. It is therefore concluded that
changing the surveillance interval from 18 months (+25%) to 24
months (+25%) will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds the current margin between the
existing Technical Specification limit and the Safety Analysis
limit. Plant equipment, which will be set at (or more conservatively
than) Technical Specification limits, will provide protective
functions to assure that Safety Analysis limits are not exceeded.
This will prevent the possibility of a new or different kind of
accident from any previously evaluated from occurring.
3. A significant reduction in a margin of safety is not
involved.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the margin which exists between the current
Technical Specification limit and the licensing basis Safety
Analysis limit. This margin, which is equivalent to the existing
margin, is necessary to assure that protective safety functions will
occur so that Safety Analysis limits are not exceeded.
[(13) Plant Vent Noble Gas Effluent Monitor (R-27):]
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Plant Vent Noble Gas Effluent Monitor (R-27) be changed from every
18 months (+25%) to every 24 months (+25%).
R-27 is a high range noble gas monitor intended for use after an
accident to provide information about the magnitude of radioactive
releases. It serves no purpose during normal operation. It provides
no function to prevent or mitigate an accident but does provide a
role in assessing the consequences of an accident. As the monitor is
a high range monitor, an estimate of the magnitude of release rather
than accuracy is important. Accordingly, continued operability of
the instrument during an extended operating cycle is more important
than the device exhibiting minimal drift characteristics.
Malfunction of the instrument would be detected by the shift checks
and functional tests performed during the extended operating cycle.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Since the monitor provides no preventive or mitigating action in
the event of an accident, no new or different type of accident has
been created by extending the operating cycle. In terms of post
accident assessment capability, alternate means exist to assess
offsite releases in the event of failure of this instrument.
3. There has been no reduction in the margin of safety.
Since the instrument provides no safety function and alternate
means exist for post accident assessment purposes, there will be no
impact on safety due to an extended period between calibrations.
[(14) Refueling Water Storage Tank Level:]
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
RWST instrumentation be changed from every 18 months (+25%) to
quarterly (once every 3 months).
A statistical analysis of channel uncertainty for a 3 month
surveillance has been performed. Based upon this analysis it has
been concluded that sufficient margin exists between the existing
Technical Specification limit and the licensing basis Safety
Analysis limit to accommodate the channel statistical error
resulting from a 3 month quarterly surveillance. The existing margin
between the Technical Specification limit and the Safety Analysis
limit provides assurance that plant protective actions will occur as
required. It is therefore concluded that changing the surveillance
interval from 18 months (+25%) to quarterly will not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed change in surveillance interval will result in a
channel statistical allowance which provides the necessary margin
between the existing Technical Specification limit and the Safety
Analysis limit. Plant equipment, which will be set at (or more
conservatively than) Technical Specification limits, will provide
protective functions to assure that Safety Analysis limits are not
exceeded. This will prevent the possibility of a new or different
kind of accident from any previously evaluated from occurring.
3. A significant reduction in a margin of safety is not
involved.
The above change in surveillance interval will result in a
channel statistical allowance which is necessary between the current
Technical Specification limit and the licensing basis Safety
Analysis limit. This margin is necessary to assure that protective
safety functions will occur so that Safety Analysis limits are not
exceeded.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for Licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Robert A. Capra.
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: March 24, 1994.
Description of Amendment Request: The changes are in support of the
forthcoming Cycle 7 for Catawba, Unit 2. The proposed Technical
Specification (TS) changes reflect:
(1) An increase from 2000 parts per million (ppm) to 2175 ppm in
the required spent fuel storage pool minimum boron concentration during
Modes 1-3 operation,
(2) An increase from 2000 ppm to 2175 ppm in the required reactor
coolant system (RCS) and refueling canal minimum boron concentration
during Mode 6 operation,
(3) The inclusion of two reload related topical reports into TS
6.9.1.9, and
(4) The revision of an administrative nature to correct errors in
nomenclature and to remove obsolete footnotes.
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:
Increase in Boron Concentration Limit for the Spent Fuel Storage Pool
(Standby Makeup Pump Water Supply)
The required spent fuel storage pool minimum boron concentration
was increased from 2000 ppm to 2175 ppm during Modes 1-3.
The proposed revision is conservative, and is required only to
maintain consistency between the boron concentration of the spent
fuel storage pool and the boron concentration of the RWST [refueling
water storage tank] during Modes 1-3 operation. Therefore, there
will be no adverse impact upon the probability or consequences of
any previously analyzed accident.
Likewise, the proposed change will not create the possibility of
a new or different kind of accident, since no new failure modes are
identified.
Finally, no negative impact upon any safety margin is created
since the proposed change is conservative.
Increase in Boron Concentration Limits for the RCS and Refueling Canal
in Mode 6
The increase in the required RCS and refueling canal minimum
boron concentration was added only to maintain consistency between
the boron concentration of the RCS and refueling canal and the RWST
in Mode 6.
The change in boron concentration limits for the RCS and
refueling canal will not increase the probability of an accident
since no accident initiators are involved with this change. Since
the change is conservative, the consequences of an accident
previously evaluated will not be increased. The increase in the
boron concentration limit for the RCS and refueling canal in Mode 6
adds further margin to the initial conditions assumed for the boron
dilution accident in the safety analysis. Therefore, the
consequences of the boron dilution accident previously evaluated
will not be increased.
The possibility of a new or different kind of accident from any
previously evaluated will not be created since this change is
bounded by previously evaluated accidents and does not introduce any
new failure modes.
This change does not involve a significant reduction in the
margin of safety since the analyses performed demonstrate that the
limits imposed meet all accident analysis and design basis
requirements.
Addition of Two Reload Related Topical Reports
This change is administrative in nature and adds two previously
approved topical reports to the list of methodologies used to
determine core operating limits. The change will have no impact upon
either the probability or consequences of a previously analyzed
accident. The methodologies described in the topical reports have
been previously reviewed and approved by the NRC. Also, no new
accident possibilities are created, since this is an administrative
change. Finally, no impact upon any safety margin is created, since
the change is administrative in nature and the described topical
reports have received full NRC approval.
Correction of Errors in Nomenclature and Removal of Obsolete Footnotes
These changes are also administrative in nature and are intended
to correct miscellaneous errors and obsolete references. As such,
the changes will have no impact upon either the probability or
consequences of any previously analyzed accidents, will not create
the possibility of any new accident scenarios, and will not impact
any safety margins.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for Licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: David B. Matthews.
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of Amendment Request: March 23, 1994.
Description of Amendment Request: The proposed amendments would
revise Technical Specification (TS) 6.9.2, ``Core Operating Limits
Report,'' to include a reference to a Duke Power Company (DPC) Topical
Report describing an analytical method for determining the core
operating limits.
Specifically, the amendments would add: ``(4) DPC-NE-1004A, Nuclear
Design Methodology Using CASMO-3/SIMULATE-3P, November 1992,'' to TS
6.9.2.
Basis for Proposed No Significant Hazards Consideration
Determination: The NRC staff reviewed Topical Report DPC-NE-1004A and
concluded in a Safety Evaluation Report dated November 23, 1992, that
the described nuclear design methodology is acceptable for performing
reload analyses for the DPC B&W 177-assembly cores in the Oconee units.
The addition of this approved nuclear design methodology to those
referenced in TS 6.9.2 provides an alternative method for determining
core operating limits such that all applicable limits (e.g., fuel
thermal mechanical limits, core thermal hydraulic limits, ECCS limits,
nuclear limits such as shutdown margin, and transient and accident
analysis limits) of the safety analysis are met. Therefore, the
proposed change to the TS (1) does not involve a significant increase
in the probability or consequences of an accident previously evaluated,
(2) does not create the possibility of a new or different kind of
accident than previously evaluated, and (3) does not involve a
significant reduction in the 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, which is
presented below:
Duke Power Company (Duke) has made the determination that this
amendment request involves a No Significant Hazards Consideration by
applying the standards established in 10 CFR 50.92. This ensures 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:
Each accident analysis addressed within the Oconee Final Safety
Analysis Report (FSAR) has been examined with respect to this
amendment request. The Technical Specifications will continue to
require operation within the bounds of the cycle-specific parameter
limits. The cycle-specific parameter limits will be calculated using
NRC approved methodology. The proposed amendment is simply an
administrative change to update the list of NRC approved methods in
Technical Specification 6.9.2. Therefore, the probability of any
Design Basis Accident (DBA) is not affected by this change, nor are
the consequences of a DBA affected by this change. This is because
the addition of an NRC approved reference to Technical Specification
6.9.2 is not considered to be an initiator or contributor to any
accident analysis addressed in the Oconee FSAR.
(2) Create the possibility of a new or different kind of
accident from any kind previously evaluated:
Operation of ONS [Oconee Nuclear Station] in accordance with
these Technical Specifications will not create any failure modes not
bounded by previously evaluated accidents. Consequently, this change
will not create the possibility of a new or different kind of
accident from any kind of accident previously evaluated.
(3) Involve a significant reduction in a margin of safety:
The Technical Specifications will continue to require operation
within the bounds of the cycle-specific parameter limits. Duke will
continue to calculate the cycle-specific parameter limits using NRC
approved methodology. In addition, each future reload will require a
10 CFR 50.59 safety review to ensure that operation of the unit
within the cycle-specific limits will not involve a reduction in a
margin of safety. Therefore, no margins of safety are affected by
the addition of an NRC approved methodology to Technical
Specification 6.9.2.
Based on the staff's analysis and its review of the licensee's
analysis, 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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
Attorney for Licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036.
NRC Project Director: David B. Matthews.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of Amendment Request: March 18, 1994.
Description of Amendment Request: The proposed amendments would
revise Technical Specification (TS) 3/4.3.3.6, Accident Monitoring
Instrumentation, TS 3/4.6.4.1, Hydrogen Monitors, and their associated
bases to incorporate the technical substance of Specification 3.3.3
from NUREG-1431, Revision O (Standard Technical Specifications) for the
Westinghouse Owners Group.
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 affect instrumentation that would be used to
assess the condition of the plant during and following an accident.
As such, the changes can have no effect on the probability of any
accident previously evaluated since this instrumentation has no
bearing on initiating events. The proposed changes will continue to
ensure the capability to monitor plant conditions during and
following an accident by requiring redundancy or diversity and
timely corrective action in the event of inoperable instrumentation.
Therefore, the proposed changes will not significantly increase the
consequences of any 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 affect the operability and action
requirements for the post accident monitoring instrumentation
system. Accordingly, the proposed changes do not involve any change
to the configuration or method of operation of any plant equipment,
and no new failure modes have been defined for any plant system or
component nor has any new limiting failure been identified as a
result of the proposed changes. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety. The intent of the existing TS requirements is
to ensure the capability to monitor the plant condition during and
following an accident so that the operators will have the
information necessary to monitor and evaluate the course of the
event and take any necessary action. Under the proposed changes this
capability will be maintained by ensuring redundancy or diversity
and by requiring timely corrective action in the event of inoperable
instrumentation. In addition, the proposed changes would avoid
unnecessary plant shutdowns by specifying an appropriate level of
action in response to inoperable instrumentation. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for Licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308.
NRC Project Director: David B. Matthews.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of Amendment Request: April 6, 1994.
Description of Amendment Request: The proposed amendment changes
the Technical Specifications to eliminate the main steam line radiation
monitor(s) (MSLRMs) reactor scram and isolation functions of the MSLRMs
currently contained in Tables 3.1.-1 and 4.1-1 of the Technical
Specifications and the associated Bases statements. This action follows
the recommendations of the BWR Owners Group (BWROG) in their Safety
Evaluation, NEDO-31400A, previously approved by the NRC Staff on May
15, 1991 by letter to the BWROG. Following is a brief description of
the proposed changes:
Tech. Spec. 3.1, ``Protective Instrumentation'' Bases is revised to
delete reference to the paragraph describing the Main Steam Line (MSL)
radiation monitoring functions for indication of excessive fuel failure
and initiation of a reactor scram and MSL isolation.
Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements
- A. Reactor Scram Functions,'' is revised to delete line Item No. 7 -
``High Radiation in Main Steam Line Tunnel.''
Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements
- B. Reactor Isolation Functions,'' is revised to delete line Item No.
6 - ``High Radiation in Main Steam Line Tunnel.''
Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements
- L. Condenser Vacuum Pump Isolation Function,'' is revised to delete
line Item No. 1 - ``High Radiation in Main Steam Line Tunnel.''
Tech. Spec. Table 4.1.1., ``Minimum Check, Calibration and Test
Frequency For Protective Instrumentation,'' is revised to delete
Instrumentation Channel No. 13 - ``High Radiation in Main Steam Line.''
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated.
The objective of the MSLRMs is to provide early indication of
gross fuel failure. The monitors provide an alarm function, and
signals that lead to a scram function and [main steam isolation
valve] MSIV isolation functions. The basis for the MSIV isolation on
an MSL high radiation signal is to reduce the quantity of fission
products transported from the reactor vessel to the condenser in the
event of gross fuel failure. No [design basis accident] DBA takes
credit for a reactor scram resulting from an MSL high radiation
signal.
The proposed change removes all trip functions of the MSLRMs.
The only modification attendant to this change is the removal of
contacts derived from the MSLRM logic to the reactor scram, reactor
isolation and offgas system isolation initiation logic. This change
does not affect the operation of any equipment having the potential
to cause a [control rod drop accident] CRDA. Therefore, the
probability of a CRDA is not increased or in any way affected by the
proposed change.
However, the CRDA analysis does take credit for MSIV isolation.
As discussed above, assuming no MSIV isolation in the event of a
CRDA, the offsite radiation doses will remain a small fraction of
the 10 CFR part 100 Reactor Site Criteria.
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 function of an MSLRM trip is to detect abnormal fission
product release and isolate the steam lines, thereby stopping the
transport of fission products from the reactor to the main
condenser. No credit is taken for the reactor scram function due to
the action of these monitors on high radiation in the MSLs in any
design basis accident. Removing the MSLRMs MSL isolation trip and
its subsequent reactor scram will not affect the operation of other
equipment or systems necessary for the prevention or mitigation of
accidents.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Eliminating the MSLRM trip functions as analyzed in NEDO-31400A
will result in a potential increase in the margin of safety because
of:
a. Improvement in the availability of the main condenser for
decay heat removal; and,
b. Elimination of inadvertent reactor scrams and challenges to
safety systems.
Therefore, operation of the facility in accordance with the
proposed changes will not result in a reduction of safety margin.
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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753.
Attorney for Licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of Amendment Request: March 9, 1994
Description of Amendment Request: The proposed amendment would
modify the Technical Specification to allow a one time exemption from
certain Appendix J testing. This exemption would extend the interval
for Type B and C testing until the Unit 2 refueling outage currently
scheduled for August 1994.
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:
As stated in 10 CFR 50.92(c), a proposed change does not involve
a significant hazards consideration if the change does not (1)
involve a significant increase in the probability or consequences of
an accident previously evaluated, or (2) the change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated, or (3) the change does not involve a
significant reduction in a margin of safety.
Criterion 1
The limiting conditions for operation involving containment
integrity are not altered by this proposed change. The surveillance
requirement concerning the Type B and C leak rate test is slightly
relaxed by the proposed change. The function of the components
affected by this surveillance are to ensure containment integrity.
Delaying the surveillance approximately two months would not change
the probability of an accident. Our significant improvement in Type
B and C leak rate test results, low anticipated leak rate for the
next surveillance, aggressive corrective actions taken, and
excellent ILRT [integrated leak rate test] results indicate there is
no reason to believe that delaying the Type B and C leak rate tests
approximately two months will cause serious deterioration to these
components. Furthermore, similar requests by utilities to extend the
surveillance beyond two years have already been found acceptable by
the NRC. Therefore, it is concluded that the proposed amendment does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2
No changes to the limiting conditions for operation for
containment integrity are proposed as part of this amendment
request. The proposed change does not involve any physical changes
to the plant or any changes to plant operations. Thus, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3
The intent of the Type B and C leak rate surveillance is to
ensure that containment integrity does not significantly
deteriorate. This is established by measuring a total leak rate of
less than 0.60 La. Our significant improvement in Type B and C
leak rate tests results, aggressive corrective actions taken, and
excellent ILRT results indicate there is no reason to believe that
delaying the Type B and C leak rate tests approximately two months
will cause serious deterioration to these components. The ``As
Found'' trend of the leak rates over the past three surveillances
indicate that the leak rate for the next surveillance will be below
the Appendix J leak rate acceptance criteria. Therefore, it is
concluded that the proposed amendment does 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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for Licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh.
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: March 23, 1994.
Description of Amendment Request: The licensee proposed to modify
Technical Specification Table 3.7-6, Area Temperature Monitoring, by
creating two zones for the main steam valve building (MSVB) and
increasing the maximum normal excursion (MNE) temperature limit for
this area from 120 deg.F to 140 deg.F.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not involve an SHC [significant hazards
consideration] because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The increase of the MNE temperature from 120 deg.F to 140
deg.F for the main steam valve building has been evaluated. The
equipment in the building has been shown to be qualified for
continuous operation at 140oF. The effect of this temperature change
has decreased slightly the qualified life of the components in the
building. For those components with a qualified life of less than 40
years, they will be replaced as a scheduled maintenance item.
An engineering review of the MSLB profile for this building was
conducted and it was concluded that those components required to
operate post accident, will continue to perform their safety
function. Therefore, since the equipment will continue to operate as
designed both during normal conditions and subsequent to a MSLB, the
probability or consequences of an accident previously evaluated is
not increased.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The effect of increasing the MNE temperature to 140 deg.F has
been evaluated and judged acceptable. The possible failure of the
equipment in this building due to the increase in temperature is no
more likely than it was before, since the equipment has been shown
to be qualified to 140 deg.F. Failure of any equipment in this
building at the new temperature will not create any new accidents or
consequences that were not considered previously.
Finally, since there are no changes in the way the plant is
operated, there is no possibility of an accident of a new or
different type than previously evaluated due to the proposed change.
3. Involve a significant reduction in margin of safety.
The proposed change increases the MNE temperature within the
MSVB. The equipment in the building has been reviewed to ensure
operability. There is a slight decrease in the qualified life, but
this was anticipated and scheduled previously and any such
replacement of equipment will continue as a maintenance item. A
review of the MSLB profile was performed for this area and it was
shown that the required equipment will continue to operate as
required.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for Licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of Amendment Request: March 15, 1994.
Description of Amendment Request: The proposed amendment would
include the use of integral fuel burnable absorbers as a method of
controlling core excess reactivity and maintaining the core power
distribution within acceptable peaking limitations.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's review is
presented below:
1. The proposed amendment would not involve a significant increase
in the probability or consequences of an accident previously evaluated.
Any fuel containing integral burnable absorbers will be analyzed
using NRC approved methods and acceptance criteria prior to being
loaded into Maine Yankee's reactor vessel core. Verification of
adequate shutdown margin is performed during low power physics testing
after each refueling. In addition, core physics monitoring is required
during power operation by Technical Specifications sections 3.10, ``CEA
Group, Power Distribution, Moderator Temperature Coefficient Limits and
Coolant Conditions,'' and 3.15 ``Reactivity Anomalies.'' Such testing
and monitoring ensures adequate margin exists to accommodate the
anticipated transients and accidents postulated in Maine Yankee's Final
Safety Analysis Report.
The licensee therefore concludes that implementation of the
proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed amendment would not create the possibility of a new
or different kind of accident from any accident previously evaluated.
A determination of compliance with approved acceptance criteria is
made for every Maine Yankee fuel reload prior to loading fuel. The use
of approved methodologies and acceptance criteria ensure that new or
different accidents will not be created by the use of integral fuel
burnable absorbers.
The licensee therefore concludes that implementation of the
proposed change will not create any or new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment would not involve a significant reduction
in a margin of safety.
The safety evaluation performed for each core reload ensures that
the core design meets appropriate acceptance criteria. Because these
criteria remain unchanged as approved by the NRC, the margin of safety
remains the same.
The licensee therefore concludes that implementation of the
proposed change would not involve a significant reduction in a margin
of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room Location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, Maine 04578.
Attorney for Licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 83 Edison Drive, Augusta, Maine 04336.
NRC Project Director: Walter R. Butler.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of Amendment Request: January 26, 1994.
Description of Amendment Request: The proposed amendment would
revise the Technical Specifications and associated Bases to reflect the
fact that the main steam isolation valves can now be tested at a
pressure of greater than or equal to Pa (42 psig) thereby
eliminating the need for the previously granted exemption to certain
Appendix J testing requirements. The exemption would no longer be
necessary because of improvements in testing technology.
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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment is limited to changes to the surveillance
testing requirements (test pressure and allowable leakage criteria)
applicable to the main steam line isolation valves. The proposed
criteria are equivalent to the current criteria with respect to
monitoring main steam isolation valve performance to ensure that
leakage past the valves would be within acceptable limits under
accident conditions. This surveillance test is performed while the
plant is in a cold shutdown condition at a time when the main steam
isolation valves are not required to be operable. Performance of the
test itself is not an input or consideration in any accident
previously evaluated, thus the proposed change will not increase the
probability of any such accident occurring.
The proposed amendment will not adversely affect the function,
operation, or reliability of the valves, nor will it diminish the
capability of the valves to perform as required during an accident.
There will be no increase in post accident off-site or on-site
radiation dose, since the adjusted leakage limit is consistent with
inputs previously established for the dose analyses. The proposed
amendment is consistent with regulatory requirements (10 CFR Part
50, Appendix J) and guidance (TER-C5257-30) that has been previously
reviewed by the NRC and found to be acceptable. Therefore, the
amendment will not increase the consequences of any accident
previously evaluated.
b. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed amendment does not involve any modification to
plant equipment or operating procedures, nor will it introduce any
new main steam isolation valve failure modes that have not been
previously considered. The proposed amendment is limited to a change
in the surveillance test pressure & acceptance criteria used to leak
test the valves. This test is performed while the plant is in a cold
shutdown condition at a time when the valves are not required to be
operable. We therefore conclude the proposed changes will not create
the possibility of a new or different kind of accident from any
accident previously analyzed.
c. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed amendment will result in the main steam isolation
valves being subjected to the maximum pressure (Pa, 42 psig)
calculated to occur under worst case accident conditions, and will
therefore provide a more realistic and challenging test of valve
performance under those conditions. The leakage rate criteria for
the test has been adjusted upward to be commensurate with the higher
test pressure, but this does not represent any increase in actual
leakage under accident conditions. On-site and off-site dose
analyses will not be affected. The proposed amendment does not
involve any change in operability requirements or limiting
conditions for operation beyond the replacement of the old test
pressure & acceptance criteria with equivalent criteria consistent
with 10 CFR Part 50, Appendix J, NUREG-1433, and TER-C5257-30. Based
on these considerations, we conclude the proposed amendment will 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for Licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh.
Philadelphia Electric Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric Company,
Docket No. 50-277, Peach Bottom Atomic Power Station, Unit No. 2, York
County, Pennsylvania
Date of Application for Amendment: April 6, 1994.
Description of Amendment Request: The amendment would reflect the
incorporation of the end-of-cycle Minimum Critical Power Ratio
Recirculation Pump Trip (MCPR-RPT) system and the replacement of the
Reactor Recirculation System (RRS) Motor Generator (M-G) Sets with
solid state adjustable speed drives (ASDs).
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 any accident previously
evaluated.
The addition of the end-of-cycle MCPR-RPT System, which utilizes
ASDs, will not have a significant increase in the probability or
consequences of an accident previously evaluated.
The end-of-cycle MCPR-RPT System has been designed to
appropriate standards and specifications to ensure that the ability
of the plant to mitigate the effects of accidents is maintained.
Additionally, the MCPR-RPT System has been analyzed such that no new
accident initiators will be created such that the probability of an
accident previously evaluated will not increase.
No new challenges to the reactor coolant pressure boundary will
result from the incorporation of the end-of-cycle MCPR-RPT System
which could result in an increase in the consequences of an
accident. All engineered safety features will function as described
in the PBAPS UFSAR [Peach Bottom Atomic Power Station Updated Final
Safety Analysis Report] in order to mitigate the consequences of
accidents previously evaluated in the PBAPS UFSAR. Additionally, all
fission product barriers and safety margins will be maintained.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The end-of-cycle MCPR-RPT System, which utilizes ASDs, has been
designed to appropriate standards and specification to ensure that
no new sequence of events or failure modes will occur such that a
transient event will escalate into a new or different type of
accident.
The software used in the digital system of the ASDs is not
subject to the verification and validation requirements discussed in
the NRC memorandum dated July 1, 1991, from A. C. Thadoni [sic]
[Thadani] (NRC) to S. A. Varga (NRC) and B. A. Bolger [sic] [Boger]
(NRC), because this equipment is neither safety-related nor
important to safety. There is no software used in the trip circuit
of the end-of-cycle MCPR-RPT System, except for the ASDs.
Additionally, the design of the modification will assure that the
new equipment EM emissions will not cause inadvertent operation of
existing plant equipment and that harmonic filters have been
incorporated to minimize electrical noise on the 13kV input power
buses.
(3) The proposed change does not result in a significant
reduction in the margin of safety.
The incorporation of the end-of-cycle MCPR-RPT System, which
utilizes ASDs, will not result in a reduction in the margin of
safety. All safety margins will be maintained.
The end-of-cycle MCPR-RPT System will aid in protecting the
integrity of the fuel barrier by tripping the recirculation pumps
early in the pressurization phase of the load rejection with no
bypass event, the turbine trip with no bypass event, and the
feedwater controller failure--maximum demand event. The early
tripping of the recirculation pumps will introduce negative void
reactivity thus reducing reactor power and maintaining safety
margins. The end-of-cycle MCPR-RPT System will ensure CPR safety
margins which protect fuel barrier integrity.
General Electric has performed a qualitative assessment of
transients that would be impacted as a result of replacing the M-G
Sets with ASDs. General Electric concluded that the faster coastdown
of the recirculation pumps during a Loss of Coolant Accident (LOCA)
due to the removal of the M-G Set inertia may slightly increase the
peak clad temperature during this event. This increase is expected
to be less than 50 deg.F. The small increase will not exceed the
2200 deg.F peak cladding temperature regulatory limit. No design or
safety limit will be exceeded.
The replacement of the M-G Sets with the ASDs will not impact
the recirculation flow controller failure--increase flow transient.
The UFSAR analysis assumes a 25%/sec rate of increase. The ASD
control system will include rate limiters that prevent a pump speed
increase greater than 25%/sec in the event of a failure. Thus, the
consequences of this transient remain bounded and safety margins
will be maintained.
The ASDs will also allow a ``soft start'' of the recirculation
pumps with the recirculation discharge valves closed prior to pump
start and a gradual increase in pump speed. This results in a
gradual change in core flow. Thus, the response to a startup of an
idle recirculation pump remains bounded by the transient analysis
and safety margins will be maintained in the transient analyses.
Changes to the fire protection equipment will still maintain the
capability to shutdown the plant in the event of a fire.
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: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: Charles L. Miller.
Philadelphia Electric Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric Company,
Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,
Units Nos. 2 and 3, York County, Pennsylvania
Date of Application for Amendments: March 28, 1994.
Description of Amendment Request: The proposed Technical
Specifications (TS) changes relocate the TS fire protection
requirements to the Updated Final Safety Analysis Report.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, 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 changes are administrative in nature and are
consistent with the guidance provided in NRC GL's [Generic Letters]
86-10 and 88-12. They do not affect the initial conditions or
precursors assumed in the Updated Final Safety Analysis Report
Section 14. These changes do not decrease the effectiveness of
equipment relied upon to mitigate the previously evaluated
accidents.
Therefore, there is no 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 previously evaluated.
The proposed changes do not make any physical changes to the
plant or changes to operating procedures. Therefore, implementation
of the proposed changes will not affect the design function or
configuration of any component or introduce any new operating
scenarios or failure modes or accident initiation.
Therefore, 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 a margin of safety.
The proposed changes are administrative in nature and are
consistent with the guidance provided in NRC GL's 86-10 and 88-12.
The proposed changes do not adversely affect the assumptions or
sequence of events used in any accident analysis.
Therefore, the proposed changes do not involve a reduction in
any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: Charles L. Miller.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of Amendment Request: January 21, 1994.
Description of Amendment Request: These amendments would revise
Technical Specifications 3.8.2.3 for both Salem Unit 1 and Salem Unit 2
to include the battery acceptance criteria, corresponding allowed
outage times and additional surveillance requirements recommended in
NUREG-1431, Standard Technical Specifications--Westinghouse Plants.
TS 3.8.2.4 ``125 Volt D.C. Distribution--Shutdown'' would also be
indirectly affected by these changes because it refers to the
surveillance requirements of TS 4.8.2.3.2 to demonstrate the battery
and chargers Operable.
In addition, Salem Unit 1 TS 3.8.2.3 Limiting Condition for
Operation (LCO) would be revised to define the specific battery charger
required for each train. Salem Unit 1 TS 3.8.2.3 Action Statement would
also be revised to restrict the use of the backup battery charger to a
period not to exceed 7 days.
Additionally, the Unit 1 action statement for an inoperable 125
volt DC bus would be modified to add the requirement that the bus also
be energized.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes do not alter plant configuration or
operation. The proposed changes do not invalidate any of the
parameters assumed in the plants UFSAR Design Basis Accident or
Transient Analyses. The proposed changes provide additional guidance
to be used to ensure operability of the safety related batteries.
New surveillance requirements and specific battery cell parameters
offer improved monitoring of the battery status. The new guidance
and surveillance requirements are consistent with the
recommendations of NUREG-1431, Standard Technical Specifications--
Westinghouse Plants, and current industry recommendations.
The changes to the Unit 1 LCO and corresponding Action Statement
restrict the use of the backup battery charger, thereby limiting the
amount of time that one AC Vital bus is allowed to power the
chargers of more than one DC train. This change brings the TS for
both Units into agreement and results in a more conservative Unit 1
TS.
Therefore, the probability or consequences of an accident
previously evaluated are not increased by the proposed change.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not introduce any design or physical
configuration changes to the facility or change the method by which
any safety-related system performs its function. The proposed
changes are consistent with the recommendations of NUREG-1431,
Standard Technical Specifications--Westinghouse Plants. Therefore,
the proposed changes will not increase the possibility of a new or
different kind of accident from any accident previously identified.
3. Does not involve a significant reduction in a margin of
safety.
The proposed changes do not alter the manner in which safety
limits or limiting safety system setpoints are determined. The new
cell parameter table and additional surveillance requirements
provide improved means to monitor and evaluate overall battery
performance. Therefore, the proposed changes do not involve a
significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
Attorney for Licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Charles L. Miller.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of Amendment Request: October 29, 1993.
Description of Amendment Request: The licensee is preparing to
replace the currently installed steam generators with new model Delta
75 steam generators (SGs). The new steam generators will be larger than
those currently installed. The physical changes to the plant and the
accident reanalyses needed to support those changes will necessitate
increasing the maximum tested charging/safety injection pump flow rate
from 680 gallons per minute to 700 gallons per minute.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of VCSNS [Virgil C. Summer Nuclear Station] in
accordance with the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Implementation of the [Delta] 75 SGs and revised operating
conditions do not contribute to the initiation of any accident
evaluated in the FSAR [Final Safety Analysis Report]. Supporting
factors are as follows:
--The [Delta] 75 SG is designed in accordance with ASME [American
Society of Mechanical Engineers] Code Section III, 1986 edition
[sic] and other applicable federal, state, and local laws, codes and
regulations and meets the original interfaces for the Model D3 SGs
with exception that provisions for a larger blowdown nozzle have
been made and the feedwater inlet nozzle is located in the upper
shell.
--All NSSS [nuclear steam supply system] components (i.e., reactor
vessel, RC Pumps, pressurizer, CRDM's [control rod drive
mechanisms], [Delta] 75 SGs, and RCS piping) are compatible with the
revised operating conditions. Their structural integrity is
maintained during all proposed plant conditions through compliance
with the ASME code.
--Fluid and auxiliary systems which are important to safety,
including the CHG/SI [charging and safety injection] system with
maximum pump flows up to 700 gpm, are not adversely impacted and
will continue to perform their design function.
--Overall plant performance and operation are not significantly
altered by the proposed changes.
Therefore, since the reactor coolant pressure boundary integrity
and system functions are not adversely impacted, the probability of
occurrence of an accident evaluated in the VCSNS FSAR will be no
greater than the original design basis of the plant.
An extensive analysis has been performed to evaluate the
consequences of the following accident types currently evaluated in
the VCSNS FSAR:
--Non-LOCA [non-loss-of-coolant accident]
--Large Break and Small Break LOCA
--Steam Generator Tube Rupture
With the [Delta] 75 SGs and revised operating conditions, the
calculated results (i.e., DNBR [departure from nucleate boiling
ratio], Primary and Secondary System Pressure, Peak Clad
Temperature, Metal Water Reaction, Challenge to Long Term Cooling,
Environmental Conditions Inside and Outside Containment, etc.) for
the accidents are similar to those currently reported in the VCSNS
FSAR. Select results (i.e., Containment Pressure during a Steam Line
Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are
slightly more limiting than those reported in the current FSAR due
to the use of the assumed operating conditions with the new [Delta]
75 SGs, and in some cases, use of an uprated core power of 2900 MWt.
However, in all cases, the calculated results do not challenge the
integrity of the primary/secondary/containment pressure boundary and
remain within the regulatory acceptance criteria applied to VCSNS's
current licensing basis. The assumptions utilized in the
radiological evaluations, described in Section 3.7, are thus
appropriate and are judged to provide a conservative estimate of the
radiological consequences during accident conditions. Given that
calculated radiological consequences are not significantly higher
than current FSAR results and remain well within 10CFR100 limits, it
is concluded that the consequences of an accident previously
evaluated in the FSAR are not increased.
(2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The [Delta] 75 SGs, revised operating conditions, and higher
allowable CHG/SI pump flows will not introduce any new accident
initiator mechanisms. Structural integrity of the RCS is maintained
during all plant conditions through compliance with the ASME code.
No new failure modes or limiting single failures have been
identified. Design requirements of auxiliary systems are met with
the RSGs [Replacement Steam Generators]. Since the safety and design
requirements continue to be met and the integrity of the reactor
coolant system pressure boundary is not challenged, no new accident
scenarios have been created. Therefore, the types of accidents
defined in the FSAR continue to represent the credible spectrum of
events to be analyzed which determine safe plant operation.
(3) The proposed license amendment does not involve a
significant reduction in a margin of safety.
Although the [Delta] 75 SGs, revised operating conditions, and
higher allowable CHG/SI pump flows will require changes to the VCSNS
Technical Specifications, it will not invalidate the LOCA, non-LOCA,
or SGTR [steam generator tube rupture] conclusions presented in the
FSAR accident analyses. For all the FSAR non-LOCA transients, the
DNB design basis, primary and secondary pressure limits, and dose
limits continue to be met. The LOCA peak cladding temperatures
remain below the limits specified in 10 CFR 50.46. The calculated
doses resulting from a SGTR event will continue to remain within a
small fraction of the 10 CFR 100 permissible releases. Environmental
conditions associated with High Energy Line Break (HELB) both inside
and outside containment have been evaluated.
The containment design pressure will not be violated as a result
of the HELB. Equipment qualification will be updated, as necessary,
to reflect the revised conditions resulting from HELB. The margin of
safety with respect to primary pressure boundary is provided, in
part, by the safety factors included in the ASME Code. Since the
components remain in compliance with the codes and standards in
effect when VCSNS was originally licensed (with the exception of the
[Delta] 75 RSGs which use the 1986 ASME Code Section III Edition),
the margin of safety is not reduced. Thus, there is no reduction in
the margin of safety as defined in the bases of the VCSNS Technical
Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180.
Attorney for Licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Project Director: William H. Bateman.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit No. 1, Lake County, Ohio
Date of Amendment Request: March 12, 1993.
Description of Amendment Request: The proposed amendment would
revise Technical Specification Table 3.3.7.1-1, to clarify the actions
to be taken if the control room ventilation radiation monitor is not
operable.
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 change clarifies Technical Specification 3.3.7.1,
``Radiation Monitoring Instrumentation'' by revising Action 72 (for an
inoperable Control Room Ventilation Radiation Monitor) to remove
several inconsistencies between it and Action 3.7.2.b.2 of the Control
Room Emergency Recirculation System Specification. Revised Action 72
simply makes the two Specifications more consistent by incorporating
alternative compensatory measures that the operators may take after the
Control Room Ventilation Radiation Monitor has been inoperable for more
than seven days. The proposed Action would retain the choice of
initiating at least one train of the Control Room Emergency
Recirculation System, while providing a second option to take which
would depend on the current Operational Condition. In Operational
Conditions 4, 5 and * * * the current Specification 3.3.7.1 Action 72
does not contain the provisions of the Control Room Emergency
Recirculation System Action 3.7.2.b.2 which directs the Operators to
suspend performance of Core Alterations, handling of irradiated fuel
and operations with a potential for draining the reactor vessel instead
of initiating the Control Room Emergency Recirculation System. This
inconsistency between the two specifications has caused compliance
difficulties; therefore, the proposed Action adds this alternative.
Also, in Operational Conditions 1, 2 and 3 a shutdown provision is
being added. The other changes are editorial, in order to clarify the
applicability of the proposed alternative compensatory measures, to be
consistent with PNPP-specific terminology, and to be more consistent
with Action b of Specification 3.7.2.
In summary, there is no change in the probability or consequences
of any accident since the revision of Specification 3.3.7.1 Action 72
is simply proposed in order to achieve consistency with the current
Action 3.7.2.b.2. Incorporation of the already approved 3.7.2.b.2
compensatory measures to suspend possible radiation accident initiating
activities provides an alternative which would actually reduce the
probability of occurrence of a previously analyzed accident, and would
have no adverse effect on accident consequences. None of the proposed
changes to the clarified action, including the editorial changes,
involves a change to the design of the plant, nor the operational
characteristics of any plant system, nor the procedures by which the
Operators run the plant.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
No design changes are being made that would create a new type of
accident or malfunction, and the methods and manner of plant operation
remains unchanged. The proposed revisions to Action 72 will remove
several inconsistencies between the two Specifications by providing
consistent actions within the Radiation Monitoring Instrumentation
Specification with those currently existing in the Control Room
Emergency Recirculation System Specification and provide an additional
shutdown requirement in Operational Conditions 1, 2 and 3. The other
changes to Action 72 are editorial, and therefore cannot affect
accident initiation parameters. The instrument to which Action 72
applies (the Control Room Ventilation Radiation Monitor (Noble Gas))
simply serves as a supporting instrumentation channel for the Control
Room Emergency Recirculation System, therefore no new or different kind
of accident can be created.
3. The proposed changes do not involve a significant reduction in a
margin of safety.
The proposed change to Specification 3.3.7.1 Action 72 simply makes
the two Specifications more consistent by making the Action for a
supporting instrumentation channel, the Control Room Ventilation
Radiation Monitor (Noble Gas), more consistent with those of the
supported system Specification, the Control Room Emergency
Recirculation System. A shutdown requirement is also being added if the
operators should choose not to initiate the supported system in
Operational Conditions 1, 2, and 3. Since the Actions of the two
Specifications will now correspond, the margin of safety as currently
exists today for the governing Specification (the Control Room
Emergency Recirculation System Specification) is maintained and the
proposed changes do not therefore reduce 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: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Attorney for Licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
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: March 18, 1994.
Description of Amendment Request: The proposed amendment would
revise TS 2.1.2 (Reactor Core), TS 2.2.1 (Reactor Protection System
Setpoints), Bases 2.1.1 and 2.1.2 (Reactor Core), Bases 2.2.1 (Reactor
Protection System Instrumentation Setpoints), TS 3.2.2 and 3.2.3 (Power
Distribution Limits), Bases 3/4.2 (Power Distribution Limits), and TS
6.9.1.7 (Administrative Controls, Core Operating Limits Report). This
amendment would remove cycle-specific limits from TS and relocate them
in the Core Operating Limits Report.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below, indicating that the proposed
changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
assumptions or probabilities are affected by the proposed relocation
of cycle-specific core operating limits to the Core Operating Limits
Report.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated. The proposed changes do not affect
any equipment, accident conditions, or assumptions which could lead
to a significant increase in radiological consequences.
2a. Not create the possibility of a new kind of accident from
any accident previously evaluated because no new accident initiators
are introduced by these proposed changes.
2b. Not create the possibility of a different kind of accident
from any accident previously evaluated because no different accident
initiators are introduced by these proposed changes.
3. Not involve a significant reduction in a margin of safety
because the proposed changes only relocate cycle-specific core
operating limits to the Core Operating Limits Report; they do not
allow less conservative operating limits. The analytical methods to
be used in the determination of cycle-specific core operating limits
are previously approved by the NRC. The same margin of safety
provided in the current Technical Specifications will continue to be
maintained.
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 Library,
Documents Department, 2801 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: John N. Hannon.
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: March 30, 1994.
Description of Amendment Request: The proposed amendment would add
a new TS Limiting Condition for Operation 3/4.4.12, Pilot Operated
Relief Valve and Block Valve, and would include associated Bases and
Surveillance Requirements.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below, indicating that the proposed
additions and changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. Automatic actuation of the PORV is not assumed
to mitigate the consequences of a design basis accident as described
in Chapter 15 of the USAR. The proposed changes will continue to
ensure the PORV and block valves are available to perform their
functions when required to do so. Therefore, it can be concluded
that the proposed changes do not involve a significant increase in
the probability of an accident previously evaluated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate accident conditions or assumptions used in evaluating the
radiological consequences of an accident.
2a. Not create the possibility of a new kind of accident from
any accident previously evaluated because the proposed changes do
not delete any function previously provided by the PORV nor has the
possibility of inadvertent opening been increased. No new types of
failures or accident initiators are introduced by the proposed
changes.
2b. Not create the possibility of a different kind of accident
from any accident previously evaluated because no new failure modes
have been defined for any plant system or component important to
safety, nor has any new limiting single failure been identified as a
result of the proposed changes. No different accident initiators or
failure mechanisms are introduced by the proposed changes.
3. Not involve a significant reduction in a margin of safety
because the proposed changes continue to ensure the availability of
the PORV and block valve when called upon to perform their function
and will not impact any safety analysis assumptions.
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 Library,
Documents Department, 2801 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: John N. Hannon.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of Amendment Request: February 14, 1994.
Brief Description of Amendments: The proposed amendment would
revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2
technical specifications to increase the Unit 2 boron concentration for
the refueling water storage tank (RWST) and the emergency core cooling
system (ECCS) accumulators to support Unit 2 operation with extended
fuel cycles. These changes are applicable to Unit 2 only and are
identical to those previously approved for Unit 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 NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. The proposed change would not increase the probability or
consequences of a previously evaluated accident.
The proposed changes are related to the boron concentration in the
RWST and ECCS accumulators. This increased concentration does not
constitute a change expected to increase the probability of a
previously evaluated accident. The means by which the proposed changes
might result in increased radiological consequences of various
accidents are discussed below.
The higher boron concentration may result in increased probability
of equipment failure following an accident due to in-containment or in-
process equipment being exposed to a more severe post-accident
environment. The general chemical properties of the slightly higher
boron concentration fluid indicates no mechanism that would result in
an appreciable increase in the component failure rate. While the
corrosive nature of the fluid will increase, this increase will be only
minimal. Thus, there is no significant increase in the consequences of
any accident due to an increase in the probability of equipment
failure.
The changes in containment spray and sump solution pH may change
the radioisotope removal and partition characteristics. While some
relevant characteristics are affected, the resulting limiting
coefficient values associated with the pH changes are bounded by the
values used in the design calculations for CPSES. Thus, no adverse
impact of the radiological consequences arising from this mechanism has
been identified.
The impact of the containment spray, with a lower pH, upon the
combustible gas production rate was also evaluated. No mechanism for
increased combustible gas production was identified.
The higher boron concentration could have an adverse impact on the
inadvertent actuation of the ECCS event. Although the timing of the
sequence of events may be affected, the departure from nucleate boiling
ratio continues to increase from its initial value throughout the
event. On the basis of its review of this event, the licensee has
identified no changes in the event probability or consequences;
however, the continued validity of this conclusion will be reconfirmed
by the licensee on a cycle-specific basis.
2. The proposed change would not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed change only changes the allowable boron concentration.
No new or different accident sequences have been identified.
Furthermore, the licensee has reviewed the heat tracing requirements
and determined that there are no additional requirements resulting from
the boron concentration increase. There are no previously unconsidered
failure mechanisms.
3. The proposed change would not involve a significant reduction in
the margin of safety.
The decrease in the containment spray and sump solution pH could be
expected to result in higher airborne iodine concentrations. The
accident source terms could be impacted by variations in the iodine
spray removal and partition factors. A comparison of the coefficients
for the minimum equilibrium containment sump solution pH to those used
in the CPSES design analyses indicated that the expected coefficient
values would remain bounded by the values used in the previous
analyses. Thus, no significant reduction in the margin of safety has
been identified.
Based on this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendments involves no significant hazards
consideration.
Local Public Document Room Location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Attorney for Licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, NW., suite 1000, Washington, DC 20036.
NRC Project Director: Suzanne C. Black.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of Amendment Request: February 14, 1994.
Brief Description of Amendments: The proposed amendment will revise
the Comanche Peak Steam Electric Station, Units 1 and 2, technical
specifications to be consistent with the new 10 CFR part 20.
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 a previously evaluated accident.
The proposed revisions to the liquid and gaseous effluent
release limits will not change the type or amount of effluent
released nor will there be an increase in individual or cumulative
dose. The changes will result in levels of radioactive materials in
effluents being maintained ALARA [as low as reasonably achievable]
and comply with 10 CFR 50.36a and 10 CFR 50 Appendix I. The change
to the high radiation area dose measurement distance will ensure
that high radiation areas are conservatively posted per 10 CFR
20.1601(a)(1) and provide controls to minimize individual dose. The
changes do not impact the operation or design of any plant
structure, system or component. Other proposed changes are
administrative only. Therefore, the proposed changes do not involve
an 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 do not affect the plant design or operation
nor do they result in a change to the configuration of any
equipment. No change is proposed that will change the type or
quantity of effluents released off site or change the source terms
available for release. Therefore, the proposed changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes do not change the type or increase the
amount of effluents released offsite. No change in the methodology
used to control radioactive waste or radiological environmental
monitoring is proposed. Control of radioactive effluents and
effluent monitor setpoints will be based on current dose to the
public limitations. Under the proposed change, high radiation area
measurements are more conservative and will not result in an
increase in individual or cumulative occupational radiation
exposures. Compliance with the limits of the revised 10 CFR 20.1301
will be demonstrated by operating within the limits of 10 CFR 50,
Appendix I and 40 CFR 190. Therefore, these changes do not reduce
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Attorney for Licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street NW., suite 1000, Washington, DC 20036.
NRC Project Director: Suzanne C. Black.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of Amendment Request: February 14, 1994.
Brief Description of Amendments: The proposed amendment would
revise the Comanche Peak Steam Electric Station Units 1 and 2 technical
specifications by reducing the frequency of reports for radiological
effluents from semiannual to annual, and change the due date from
within 60 days after January 1 and July 1 to prior to May 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:
1. The proposed changes do not involve a significant increase in
the probability or consequences of a previously evaluated accident.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The amendment involves only changes of reporting
frequency and due date requirements for radiological effluent
release reporting. These changes are administrative in nature and do
not affect safe operation of the plant; therefore, accident
probabilities or consequences are unaffected.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed amendment is administrative in nature and
does not involve any changes to plant design of configuration. For
this reason, it will not create the possibility of a new or
different kind of accident.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed amendment does not involve a significant reduction
in the margin of safety. The proposed amendment only changes the
reporting frequency and due date requirements for radiological
effluent release reporting. The reporting requirements for
radiological effluent releases are administrative changes:
therefore, there is not 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: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Attorney for Licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street NW., suite 1000, Washington, DC 20036.
NRC Project Director: Suzanne C. Black.
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: March 30, 1994.
Description of Amendment Request: The proposed changes would revise
the North Anna Power Station, Units No. 1 and No. 2 (NA-1&2) Technical
Specifications (TS). Specifically, the proposed changes would revise
the High Head Safety Injection (HHSI) flow balance surveillance
requirements by removing specific numerical values. The numerical
values would be replaced with broader requirements to ensure that the
HHSI flow rates meet the loss of coolant accident (LOCA) analysis
acceptance criteria and pump runout limits. The NA-1&2 TS 4.5.2.h
requires a surveillance test of the HHSI system following the
completion of any modification to the Emergency Core Cooling System
(ECCS) subsystems that could alter the subsystem flow characteristics.
The current surveillance criteria specify values for the sum of the
injection line flow rates, excluding the highest flow rate, and the
total pump flow rate. These correspond to requirements for the safety
analysis flow input and the HHSI pump runout limit, respectively.
The HHSI test acceptance criteria in the current TS are very narrow
because of the various system physical and technical constraints that
need to be considered in the flow balance testing. These acceptance
criteria may also be more restrictive than required by either the LOCA
analysis or the actual pump runout requirements. For example, the LOCA
analysis contains input conservatisms that could be used to offset a
reduction in the required HHSI flow while still meeting the 10 CFR
50.46 LOCA acceptance criteria. The proposed TS changes would permit
the use of additional available margin, while maintaining a strong
technical linkage between the measured system performance and the
safety analysis. Although these proposed TS changes remove the
numerical values from TS 4.5.2.h, neither the methodology nor the
acceptance criteria for LOCA analysis are affected.
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:
Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specification changes will
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
Technical Specification changes continue to require that with one
HHSI pump running, the sum of the flows through the two lowest
branch lines shall be [greater than or equal to] the minimum HHSI
flow required by the safety analysis and that the total HHSI pump
flow rate shall be [less than or equal to] the evaluated HHSI pump
runout limit.
Likewise, the consequences of the accidents previously evaluated
will not increase as a result of the proposed Technical
Specification changes. The system performance will remain bounded by
the safety analysis for all postulated conditions. The safety
analysis will continue to be performed and evaluated in accordance
with the requirements of 10 CFR 50.59 and 10 CFR 50.46.
2. Create the possibility of a new or different kind of accident
or malfunction from any previously evaluated. The proposed Technical
Specification changes will not affect the capability of the HHSI
System to perform its intended function. The proposed Technical
Specification changes are bounded by the existing safety analysis
and do not involve operation of plant equipment in a different
manner from which it was designed to operate.
Since a new failure mode is not created, a new or different type
of accident or malfunction is not created.
3. Involve a reduction in a margin of safety. The system
performance will continue to bound the flow rates specified in the
safety analysis, therefore safety margins are not reduced.
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: Herbert N. Berkow.
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.
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, STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of Application for Amendments: March 21, 1994.
Description of Amendment Requests: The proposed amendments would
permit continued activities at all four units with main steam Code
safety valve tolerances of plus or minus 3% until the lift setpoints
can be reset to within plus or minus 1%.
Date of Publication of Individual Notice in Federal Register: March
29, 1994 (59 FR 14685).
Expiration Date of Individual Notice: April 29, 1994.
Local Public Document Room Location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of Amendment Request: March 24, 1994.
Description of Amendment Request: The proposed amendment would
revise section 6.0 (Administrative Controls). Specifically, the plant
staff requirement (specified in Technical Specification (TS) 6.2.2.i)
would be revised to temporarily allow the operations manager to have
held a senior reactor operator (SRO) license at a pressurized water
reactor (PWR) other than Indian Point 3. The TS currently requires the
operations manager to have or have held an SRO license at Indian Point
3 only. This proposed change is needed to allow management changes at
the facility in an effort to improve overall performance. The proposed
changes would be in effect for a period ending 3 years after restart
from the 1993/1994 Performance Improvement Outage.
Date of Publication of Individual Notice in Federal Register: April
1, 1994 (59 FR 15464).
Expiration Date of Individual Notice: May 3, 1994.
Local Public Document Room Location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for Licensee: Charles M. Pratt, 10 Columbus Circle, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of Application for Amendments: December 2, 1993.
Brief Description of Amendments: The amendments will modify TS 3/
4.6.1.2 by removing the schedular requirements for a Type A (overall
integrated containment leakage rate) test to be performed specifically
at 40 plus or minus 10-month intervals and replacing these requirements
with a requirement to perform Type A testing in accordance with
Appendix J to 10 CFR part 50.
Date of Issuance: April 6, 1994.
Effective date: April 6, 1994.
Amendment Nos.: 73, 59, and 45.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of Initial Notice in Federal Register: January 5, 1994 (59 FR
616) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 6, 1994.
No significant Hazards Consideration Comments Received: No.
Local Public Document Room Location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of Application for Amendments: August 27, 1993, as
supplemented March 11, 1994.
Brief Description of Amendments: The amendments revise the Calvert
Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specifications
(TSs) by removing the list of containment isolation valves in Table
3.6-1. The amendments also make accompanying changes to various TSs and
to the TS Bases. These amendments are a ``line-item'' TS improvement
and follow the guidance of Generic Letter 91-08, ``Removal of Component
Lists From Technical Specifications.''
Date of issuance: April 7, 1994.
Effective Date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 187 and 164
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: September 29, 1993 (58
FR 50966) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 7, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Calvert County Library, Prince
Frederick, Maryland 20678.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of Application for Amendment: June 7, 1993, August 9, and
December 10, 1993.
Brief Description of Amendment: This amendment revises the
Technical Specification (TS) to support a 24-month fuel cycle. The TS
changes include extending surveillance intervals and adjusting
setpoints as justified in the Safety Evaluation.
Date of Issuance: April 6, 1994.
Effective Date: April 6, 1994.
Amendment No.: 151.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of Initial Notice in Federal Register: March 16, 1994 (59 FR
2863) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 6, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of Application for Amendments: March 26, 1993.
Brief Description of Amendments: The amendments modify the trip
level settings for the Isolation Condenser and High Pressure Core
Injection System Steam lines to more conservative values. In addition,
the proposed amendments revise the ECCS Low-Low Water Level initiation
trip setting to a more conservative number.
Date of Issuance: April 5, 1994.
Effective Date: April 5, 1994.
Amendment Nos.: 126 and 120.
Facility Operating License Nos. DPR-19 and DPR-25. The amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: March 2, 1994 (59 FR
10002) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 5, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Morris Public Library, 604
Liberty Street, Morris, Illinois 60450.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of Application for Amendments: January 24, 1994.
Brief Description of Amendments: The amendments implement line item
5.9 of Generic Letter 93-05, ``Line-Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation'', which provided recommendations for deleting the
requirement to perform response time testing where the required time
corresponds to the diesel start time.
Date of issuance: April 7, 1994.
Effective Date: April 7, 1994.
Amendment Nos.: 98 and 82.
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: February 16, 1994 (59
FR 7686). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 7, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of Application for Amendments: February 22, 1993 as
supplemented August 16, 1993.
Brief Description of Amendments: The amendments allow continued
operation of one unit for a period of seven days while the common plant
(Division 1) emergency diesel generator (``O'' DG) is out of service
for the performance of specified Technical Specification surveillance
requirements and the performance of planned maintenance and/or
modification work. Also, the amendments clarify Surveillance
Requirement 4.8.1.1.2.a.7 to allow an emergency diesel generator to
remain Operable with only one air start subsystem pressurized.
Date of Issuance: April 11, 1994.
Effective Date: April 11, 1994.
Amendment Nos.: 99 and 83.
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: July 7, 1993 (58 FR
36430) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 11, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
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 28, 1993, as
supplemented February 17, 1994.
Brief Description of Amendments: The amendments delete the portion
of the 18-month surveillance requirement contained in Technical
Specification (TS) 4.5.2.d associated with verifying that the decay
heat removal system suction isolation valves automatically close on a
reactor coolant system pressure signal. Also, an obsolete footnote to
TS 4.5.2.e is being deleted. This footnote is no longer necessary since
the first Unit 1 refueling outage is complete.
Date of Issuance: April 4, 1994.
Effective Date: April 4, 1994.
Amendment Nos.: 117 and 111.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: March 2, 1994 (59 FR
10004) The February 17, 1994, letter provided clarifying information
that did not change the scope of the initial September 28, 1993,
application and initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of Application for Amendment: July 22, 1993, as supplemented
by letter dated October 20, 1993.
Brief Description of Amendment: The amendment removed the cycle-
specific variables from the Technical Specifications (TSs) and
controlled them under a new document called the Core Operating Limits
Report (COLR), in accordance with Generic Letter 88-16.
Date of Issuance: April 11, 1994.
Effective Date: April 11, 1994.
Amendment No.: 157.
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of Initial Notice in Federal Register: September 1, 1993 (58
FR 46230). The additional information contained in the supplemental
letter dated October 20, 1993, was clarifying in nature and, thus,
within the scope of the initial notice and did not affect the staff's
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 11, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power & Light
Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1,
Claiborne County, Mississippi
Date of Application for Amendment: January 13, 1994.
Brief Description of Amendment: The amendment requested the removal
of the temporary technical specification limit on the number of spent
fuel assemblies that may be stored in the spent fuel pool at Grand Gulf
Nuclear Station pending licensee verification of the adequacy of the
spent fuel pool heat removal capability.
Date of Issuance: April 4, 1994.
Effective Date: April 4, 1994.
Amendment No: 113.
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of Initial Notice in Federal Register: March 2, 1994 (59 FR
10006) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 4, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Judge George W. Armstrong
Library, Post Office Box 1406, S. Commerce at Washington, Natchez,
Mississippi 39120.
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: April 20, 1993.
Brief Description of Amendments: These amendments delete the lead/
lag compensator term on the measured reactor coolant system loop
temperature difference from the overtemperature and overpower Delta T
reactor trip functions.
Date of Issuance: April 4, 1994.
Effective Date: April 4, 1994.
Amendment Nos. 161 and 155.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: June 9, 1993 (58 FR
32383) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 4, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Florida International
University, University Park, Miami, Florida 33199.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,
Georgia
Date of Application for Amendments: September 20, 1993.
Brief Description of Amendments: The amendments revise the Units 1
and 2 Channel Functional Test frequency from quarterly to once per 18
months for the scram discharge volume float type level switches.
Date of Issuance: April 15, 1994.
Effective Date: To be implemented within 60 days from the date of
issuance.
Amendment Nos.: 193 and 133.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: October 27, 1993 (58 FR
57852) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 15, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
No. 50-499, South Texas Project, Unit 2, Matagorda County, Texas
Date of Amendment Request: January 25, 1994.
Brief Description of Amendment: The amendment added new Technical
Specifications, 3/4.10.6 and 3/4.10.7, to the Special Test Exceptions
section. TS 3/4.10.6 allows the restart of Unit 2 with expired
calibrations on the core exit thermocouples (CET) and the reactor
coolant system (RCS) resistance temperature detectors (RTD) by setting
aside the affected limiting conditions for operation (LCOs) until the
calibrations are complete. This is a one-time only change that is valid
during the third refueling outage for Unit 2 until the calibrations are
complete. TS 3/4.10.7 adds a new technical specification to allow the
ascension to 75 percent rated thermal power with an expired precision
heat balance reactor coolant flow measurement. This change is effective
only for Unit 2, Cycle 4, until the surveillance requirement is
completed.
Date of Issuance: April 1, 1994.
Effective Date: April 1, 1994, to be implemented within 10 days of
issuance.
Amendment No.: Amendment No. 48.
Facility Operating License No. NPF-80. Amendment revised the
Technical Specifications.
Date of Initial Notice in Federal Register: February 16, 1994 (59
FR 7690) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 1, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of Application for Amendment: January 21, 1994.
Brief Description of Amendment: The amendment revises Technical
Specification 4.6.3 (Emergency Power Sources), to eliminate unnecessary
testing of an operable emergency diesel generator (EDG) when the
redundant EDG becomes inoperable. This amendment is intended to
increase EDG reliability and the overall level of plant safety by
reducing the stresses on the EDGs caused by unnecessary testing. This
amendment also eliminates the requirement to load the operable EDG with
the offsite network when it is tested with one EDG inoperable.
Date of Issuance: April 6, 1994.
Effective Date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 147.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of Initial Notice in Federal Register: March 2, 1994 (59 FR
10009) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 6, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of Application for Amendment: November 18, 1993.
Brief Description of Amendment: The amendment revises the setpoints
for the degraded voltage relays for the 4.16kV Power Boards 102 and 103
as specified in Technical Specification Table 3.6.2i. The setpoints
have been revised from 3580 volts 3 seconds to 3705 volts
> 3.4 seconds and < 60="" seconds.="" this="" change="" has="" been="" made="" in="" response="" to="" findings="" of="" the="" nrc's="" electrical="" distribution="" system="" functional="" inspection="" conducted="" at="" nine="" mile="" point="" nuclear="" station="" unit="" no.="" 1="" from="" september="" 23,="" 1991,="" to="" october="" 25,="" 1991.="" date="" of="" issuance:="" april="" 7,="" 1994.="" effective="" date:="" as="" of="" the="" date="" of="" issuance="" to="" be="" implemented="" prior="" to="" startup="" from="" the="" next="" refueling="" outage.="" amendment="" no.:="" 148.="" facility="" operating="" license="" no.="" dpr-63:="" amendment="" revises="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" december="" 22,="" 1993="" (58="" fr="" 67851).="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 7,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" reference="" and="" documents="" department,="" penfield="" library,="" state="" university="" of="" new="" york,="" oswego,="" new="" york="" 13126.="" north="" atlantic="" energy="" service="" corporation,="" docket="" no.="" 50-443,="" seabrook="" station,="" unit="" no.="" 1,="" rockingham="" county,="" new="" hampshire="" date="" of="" amendment="" request:="" october="" 28,="" 1993.="" description="" of="" amendment="" request:="" the="" amendment="" implements="" 13="" of="" 47="" line="" item="" technical="" specification="" (ts)="" improvements="" recommended="" by="" generic="" letter="" 93-05.="" most="" of="" the="" changes="" revise="" the="" allowable="" time="" intervals="" for="" performing="" certain="" surveillance="" requirements="" (sr)="" on="" various="" plant="" components="" during="" power="" operation="" or="" delete="" the="" requirement="" entirely="" or="" under="" certain="" conditions.="" one="" change="" modifies="" testing="" requirements="" identified="" in="" an="" action="" statement.="" specifically,="" the="" amendment="" modifies="" surveillance="" requirements="" 4.1.3.1.2,="" 4.6.4.1,="" 4.3.2.1="" (table="" 4.3-2,="" functional="" unit="" 3.c.4),="" 4.3.3.1="" (table="" 4.3-3,="" functional="" units="" 1="" through="" 6),="" 4.4.6.2.2,="" 4.4.11.1,="" 4.4.3.2,="" 4.5.1.1.1,="" 4.5.1.1.2,="" 4.5.2,="" 4.6.2.1,="" 4.6.4.2,="" 4.7.1.2.1,="" and="" the="" action="" statements="" in="" technical="" specification="" 3.8.1.1.="" date="" of="" issuance:="" april="" 7,="" 1994.="" effective="" date:="" as="" of="" the="" date="" of="" issuance="" to="" be="" implemented="" within="" 60="" days.="" amendment="" no.:="" 30.="" facility="" operating="" license="" no.="" npf-86:="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" february="" 2,="" 1994="" (59="" fr="" 4942).="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 7,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" exeter="" public="" library,="" 47="" front="" street,="" exeter,="" new="" hampshire="" 03833.="" northern="" states="" power="" company,="" docket="" no.="" 50-263,="" monticello="" nuclear="" generating="" plant,="" wright="" county,="" minnesota="" date="" of="" application="" for="" amendment:="" july="" 7,="" 1993.="" brief="" description="" of="" amendment:="" the="" amendment="" changes="" technical="" specification="" 3.6.d,="" ``primary="" system="" boundary,="" coolant="" leakage,''="" and="" the="" corresponding="" surveillance="" requirements.="" the="" amendment="" adds="" a="" clause="" to="" make="" the="" operability="" requirements="" of="" leakage="" measurement="" instruments="" applicable="" only="" when="" irradiated="" fuel="" is="" in="" the="" reactor="" and="" reactor="" water="" temperature="" is="" above="" 212="" deg.f.="" with="" regards="" to="" leakage="" measurement="" instruments,="" it="" is="" now="" required="" that="" leak="" rate="" measurements="" be="" made="" once="" per="" 12="" hours.="" in="" addition,="" instruments="" must="" be="" restored="" to="" operable="" status="" within="" 30="" days="" or="" else="" shutdown="" would="" be="" required.="" operability="" requirements="" for="" the="" drywell="" particulate="" radioactivity="" monitoring="" system="" are="" now="" addressed.="" surveillance="" requirements="" regarding="" primary="" containment="" atmosphere,="" identified="" and="" unidentified="" leakage="" of="" reactor="" coolant,="" and="" performance="" of="" a="" sensor="" check="" for="" the="" primary="" containment="" sump="" leakage="" measurement="" system="" are="" changed="" to="" once="" per="" shift,="" not="" to="" exceed="" 12="" hours.="" date="" of="" issuance:="" april="" 15,="" 1994.="" effective="" date:="" april="" 15,="" 1994.="" amendment="" no.:="" 87.="" facility="" operating="" license="" no.="" dpr-22.="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" august="" 4,="" 1993="" (58="" fr="" 41507)="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 15,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" minneapolis="" public="" library,="" technology="" and="" science="" department,="" 300="" nicollet="" mall,="" minneapolis,="" minnesota="" 55401.="" pennsylvania="" power="" and="" light="" company,="" docket="" no.="" 50-388,="" susquehanna="" steam="" electric="" station,="" unit="" 2,="" luzerne="" county,="" pennsylvania="" date="" of="" application="" for="" amendment:="" november="" 24,="" 1993,="" and="" supplemented="" by="" letters="" dated="" january="" 7,="" and="" february="" 14,="" 1994.="" brief="" description="" of="" amendment:="" the="" amendment="" raised="" the="" authorized="" power="" level="" from="" the="" 3293="" mwt="" to="" a="" new="" limit="" of="" 3441="" mwt.="" the="" amendment="" also="" changed="" the="" technical="" specifications="" to="" implement="" uprated="" power="" operation.="" date="" of="" issuance:="" april="" 11,="" 1994.="" effective="" date:="" as="" of="" its="" date="" of="" issuance="" and="" is="" to="" be="" implemented="" prior="" to="" startup="" in="" cycle="" 7,="" currently="" scheduled="" to="" occur="" may="" 21,="" 1994.="" amendment="" no.:="" 103.="" facility="" operating="" license="" no.="" npf-22.="" this="" amendment="" revised="" the="" technical="" specifications="" and="" the="" license.="" date="" of="" initial="" notice="" in="" federal="" register:="" december="" 22,="" 1993="" (58="" fr="" 67852)="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 11,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" osterhout="" free="" library,="" reference="" department,="" 71="" south="" franklin="" street,="" wilkes-barre,="" pennsylvania="" 18701.="" pennsylvania="" power="" and="" light="" company,="" docket="" nos.="" 50-387="" and="" 50-388="" susquehanna="" steam="" electric="" station,="" units="" 1="" and="" 2,="" luzerne="" county,="" pennsylvania.="" date="" of="" application="" for="" amendments:="" april="" 16,="" 1993.="" brief="" description="" of="" amendments:="" the="" amendments="" revised="" the="" technical="" specifications="" to="" conform="" to="" the="" nrc="" staff="" positions="" on="" inservice="" inspection="" and="" on="" monitoring="" of="" unidentified="" leakage="" in="" generic="" letter="" 88-01,="" ``nrc="" position="" on="" intergranular="" stress="" corrosion="" cracking="" in="" bwr="" austenitic="" stainless="" steel="" piping''.="" date="" of="" issuance:="" april="" 15,="" 1994.="" effective="" date:="" april="" 15,="" 1994.="" amendment="" nos.:="" 134="" and="" 104.="" facility="" operating="" license="" nos.="" npf-14="" and="" npf-22.="" these="" amendments="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" may="" 12,="" 1993="" (58="" fr="" 28058).="" the="" commission's="" related="" evaluation="" of="" the="" amendments="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 15,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" osterhout="" free="" library,="" reference="" department,="" 71="" south="" franklin="" street,="" wilkes-barre,="" pennsylvania="" 18701.="" philadelphia="" electric="" company,="" public="" service="" electric="" and="" gas="" company,="" delmarva="" power="" and="" light="" company,="" and="" atlantic="" city="" electric="" company,="" docket="" nos.="" 50-277="" and="" 50-278,="" peach="" bottom="" atomic="" power="" station,="" unit="" nos.="" 2="" and="" 3,="" york="" county,="" pennsylvania="" date="" of="" application="" for="" amendments:="" may="" 21,="" 1992="" brief="" description="" of="" amendments:="" the="" amendments="" revised="" the="" expiration="" dates="" from="" january="" 31,="" 2008,="" for="" units="" 2="" and="" 3,="" to="" august="" 8,="" 2013,="" for="" unit="" 2,="" and="" july="" 2,="" 2014,="" for="" unit="" 3.="" the="" original="" expiration="" date="" is="" 40="" years="" from="" the="" date="" of="" issuance="" of="" the="" construction="" permit="" for="" both="" units.="" the="" revised="" dates="" are="" 40="" years="" from="" the="" date="" of="" issuance="" of="" the="" respective="" operating="" licenses="" (i.e.,="" august="" 8,="" 1973="" for="" unit="" 2="" and="" july="" 2,="" 1974="" for="" unit="" 3).="" date="" of="" issuance:="" march="" 28,="" 1994.="" effective="" date:="" march="" 28,="" 1994.="" amendments="" nos.:="" 186="" and="" 191.="" facility="" operating="" license="" nos.="" dpr-44="" and="" dpr-56:="" amendments="" revised="" the="" licenses.="" date="" of="" initial="" notice="" in="" federal="" register:="" august="" 5,="" 1992="" (57="" fr="" 34590).="" the="" commission's="" related="" evaluation="" of="" the="" amendments="" is="" contained="" in="" a="" safety="" evaluation="" dated="" march="" 28,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" government="" publications="" section,="" state="" library="" of="" pennsylvania="" (regional="" depository),="" education="" building,="" walnut="" street="" and="" commonwealth="" avenue,="" box="" 1601,="" harrisburg,="" pennsylvania="" 17105.="" philadelphia="" electric="" company,="" public="" service="" electric="" and="" gas="" company,="" delmarva="" power="" and="" light="" company,="" and="" atlantic="" city="" electric="" company,="" docket="" nos.="" 50-277="" and="" 50-278,="" peach="" bottom="" atomic="" power="" station,="" unit="" nos.="" 2="" and="" 3,="" york="" county,="" pennsylvania="" date="" of="" application="" for="" amendments:="" november="" 17,="" 1993.="" brief="" description="" of="" amendments:="" these="" amendments="" revise="" the="" surveillance="" requirements="" to="" eliminate="" unnecessary="" diesel="" generator="" testing="" when="" a="" diesel="" generator="" or="" an="" offsite="" power="" source="" becomes="" inoperable.="" this="" change="" reduces="" the="" stresses="" on="" the="" diesel="" generators="" caused="" by="" unnecessary="" testing.="" date="" of="" issuance:="" april="" 5,="" 1994.="" effective="" date:="" april="" 5,="" 1994.="" amendments="" nos.:="" 187="" and="" 192.="" facility="" operating="" license="" nos.="" dpr-44="" and="" dpr-56:="" amendments="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" january="" 5,="" 1994="" (59="" fr="" 628).="" the="" commission's="" related="" evaluation="" of="" the="" amendments="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 5,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" government="" publications="" section,="" state="" library="" of="" pennsylvania="" (regional="" depository),="" education="" building,="" walnut="" street="" and="" commonwealth="" avenue,="" box="" 1601,="" harrisburg,="" pennsylvania="" 17105.="" philadelphia="" electric="" company,="" public="" service="" electric="" and="" gas="" company,="" delmarva="" power="" and="" light="" company,="" and="" atlantic="" city="" electric="" company,="" docket="" nos.="" 50-277="" and="" 50-278,="" peach="" bottom="" atomic="" power="" station,="" unit="" nos.="" 2="" and="" 3,="" york="" county,="" pennsylvania="" date="" of="" application="" for="" amendments:="" may="" 25,="" 1993,="" as="" supplemented="" march="" 11,="" 1994.="" brief="" description="" of="" amendments:="" these="" administrative="" amendments="" (1)="" remove="" references="" to="" the="" service="" platform="" hoist,="" (2)="" correct="" a="" typographical="" error="" concerning="" the="" emergency="" transformer="" degraded="" voltage="" relay="" setpoint="" tolerance,="" and="" (3)="" clarify="" that="" the="" basis="" for="" recalibration="" of="" certain="" pressure="" switches="" is="" reactor="" thermal="" power="" instead="" of="" turbine="" first="" stage="" pressure.="" date="" of="" issuance:="" april="" 7,="" 1994.="" effective="" date:="" april="" 7,="" 1994.="" amendments="" nos.:="" 188="" and="" 193.="" facility="" operating="" license="" nos.="" dpr-44="" and="" dpr-56:="" amendments="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" july="" 21,="" 1993="" (58="" fr="" 39059).="" the="" commission's="" related="" evaluation="" of="" the="" amendments="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 7,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" government="" publications="" section,="" state="" library="" of="" pennsylvania="" (regional="" depository),="" education="" building,="" walnut="" street="" and="" commonwealth="" avenue,="" box="" 1601,="" harrisburg,="" pennsylvania="" 17105.="" philadelphia="" electric="" company,="" public="" service="" electric="" and="" gas="" company="" delmarva="" power="" and="" light="" company,="" and="" atlantic="" city="" electric="" company,="" docket="" nos.="" 50-277="" and="" 50-278,="" peach="" bottom="" atomic="" power="" station,="" unit="" nos.="" 2="" and="" 3,="" york="" county,="" pennsylvania="" date="" of="" application="" for="" amendments:="" december="" 21,="" 1993,="" as="" supplemented="" on="" march="" 11,="" 1994.="" brief="" description="" of="" amendments:="" these="" amendments="" revise="" technical="" specification="" (ts)="" table="" 3.2.f="" to="" accurately="" describe="" the="" main="" stack="" high="" range="" and="" reactor="" building="" roof="" vent="" high="" range="" radiation="" monitors,="" and="" delete="" previously="" approved="" amendment="" no.="" 168="" for="" unit="" 3.="" amendment="" no.="" 168="" was="" an="" emergency="" temporary="" change="" which="" is="" no="" longer="" requested.="" date="" of="" issuance:="" april="" 7,="" 1994.="" effective="" date:="" april="" 7,="" 1994.="" amendments="" nos.:="" 189="" and="" 194.="" facility="" operating="" license="" nos.="" dpr-44="" and="" dpr-56:="" amendments="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" february="" 16,="" 1994="" (59="" fr="" 7697)="" the="" commission's="" related="" evaluation="" of="" the="" amendments="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 7,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" government="" publications="" section,="" state="" library="" of="" pennsylvania,="" (regional="" depository)="" education="" building,="" walnut="" street="" and="" commonwealth="" avenue,="" box="" 1601,="" harrisburg,="" pennsylvania="" 17105.="" power="" authority="" of="" the="" state="" of="" new="" york,="" docket="" no.="" 50-333,="" james="" a.="" fitzpatrick="" nuclear="" power="" plant,="" oswego="" county,="" new="" york="" date="" of="" application="" for="" amendment:="" december="" 28,="" 1993.="" brief="" description="" of="" amendment:="" the="" amendment="" clarifies="" limiting="" condition="" for="" operation="" (lco)="" 3.5.d.4.="" amendment="" no.="" 179="" to="" the="" ts="" added="" lco="" 3.5.d.4="" to="" permit="" hydrostatic="" and="" leakage="" testing="" at="" temperatures="" up="" to="" 300="" deg.f="" without="" requiring="" certain="" equipment,="" including="" the="" automatic="" depressurization="" system="" (ads),="" to="" be="" operable.="" however,="" lco="" 3.5.d.4="" can="" be="" mistakenly="" interpreted="" to="" require="" the="" ads="" be="" operable="" at="" temperatures="" less="" than="" 212="" deg.f.="" requiring="" the="" ads="" to="" be="" operable="" during="" hydrostatic="" and="" leakage="" testing="" with="" temperatures="" below="" 212="" deg.f="" was="" clearly="" not="" the="" intent="" of="" amendment="" no.="" 179.="" the="" amendment="" clarifies="" lco="" 3.5.d.4="" to="" resolve="" this="" concern="" and="" is="" considered="" an="" administrative="" change.="" date="" of="" issuance:="" april="" 6,="" 1994.="" effective="" date:="" as="" of="" the="" date="" of="" issuance="" to="" be="" implemented="" within="" 30="" days.="" amendment="" no.:="" 209.="" facility="" operating="" license="" no.="" dpr-59:="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" march="" 2,="" 1994="" (59="" fr="" 10014)="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 6,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" reference="" and="" documents="" department,="" penfield="" library,="" state="" university="" of="" new="" york,="" oswego,="" new="" york="" 13126.="" power="" authority="" of="" the="" state="" of="" new="" york,="" docket="" no.="" 50-333,="" james="" a.="" fitzpatrick="" nuclear="" power="" plant,="" oswego="" county,="" new="" york="" date="" of="" application="" for="" amendment:="" december="" 29,="" 1993.="" brief="" description="" of="" amendment:="" the="" technical="" specifications="" (tss)="" amendment="" revised="" section="" 3.6.d.4="" to="" eliminate="" an="" inconsistency="" between="" the="" operability="" requirements="" for="" the="" reactor="" coolant="" system="" (rcs)="" leakage="" detection="" and="" the="" specified="" requirements="" for="" monitoring="" rcs="" leakage.="" additionally,="" the="" amendment="" revised="" the="" tss="" to="" make="" numerous="" editorial="" corrections="" which="" are="" administrative="" in="" nature.="" date="" of="" issuance:="" april="" 13,="" 1994.="" effective="" date:="" as="" of="" the="" date="" of="" issuance="" to="" be="" implemented="" within="" 30="" days.="" amendment="" no.:="" 210.="" facility="" operating="" license="" no.="" dpr-59:="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" february="" 2,="" 1994="" (59="" fr="" 4945)="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 13,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" reference="" and="" documents="" department,="" penfield="" library,="" state="" university="" of="" new="" york,="" oswego,="" new="" york="" 13126.="" public="" service="" electric="" &="" gas="" company,="" docket="" no.="" 50-354,="" hope="" creek="" generating="" station,="" salem="" county,="" new="" jersey="" date="" of="" application="" for="" amendment:="" august="" 30,="" 1993,="" and="" supplement="" dated="" march="" 21,="" 1994.="" brief="" description="" of="" amendment:="" the="" amendment="" revises="" the="" composition="" of="" the="" station="" operations="" review="" committee="" (sorc)="" and="" increases="" the="" submittal="" interval="" of="" the="" radiological="" effluent="" release="" report="" from="" semiannually="" to="" annually.="" date="" of="" issuance:="" april="" 15,="" 1994.="" effective="" date:="" april="" 15,="" 1994.="" amendment="" no.:="" 67.="" facility="" operating="" license="" no.="" npf-57:="" this="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" september="" 29,="" 1993="" (58="" fr="" 50973)="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 15,="" 1994.="" 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:="" april="" 23,="" 1993,="" and="" supplemented="" by="" letters="" dated="" november="" 10,="" 1993="" and="" january="" 13,="" 1994.="" brief="" description="" of="" amendment:="" the="" amendment="" lowers="" the="" technical="" specification="" limit="" for="" the="" maximum="" ultimate="" heat="" sink="" temperature="" and="" revise="" the="" bases="" for="" the="" station="" service="" water="" system.="" date="" of="" issuance:="" april="" 15,="" 1994.="" effective="" date:="" april="" 15,="" 1994.="" amendment="" no.:="" 68.="" facility="" operating="" license="" no.="" npf-57:="" this="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 15,="" 1994.="" 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:="" december="" 8,="" 1993.="" brief="" description="" of="" amendments:="" these="" amendments="" incorporate="" the="" guidance="" of="" nrc="" generic="" letter="" 90-06="" that="" addresses="" power-operated="" relief="" valve="" and="" block="" valve="" reliability="" and="" additional="" low-temperature="" overpressure="" protection="" for="" light="" water="" reactors.="" date="" of="" issuance:="" april="" 7,="" 1994.="" effective="" date:="" april="" 7,="" 1994.="" amendment="" nos.="" 150="" and="" 130.="" facility="" operating="" license="" nos.="" dpr-70="" and="" dpr-75:="" these="" amendments="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" january="" 19,="" 1994="" (59="" fr="" 2870)="" the="" commission's="" related="" evaluation="" of="" the="" amendments="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 7,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" salem="" free="" public="" library,="" 112="" west="" broadway,="" salem,="" new="" jersey="" 08079="" southern="" nuclear="" operating="" company,="" inc.,="" docket="" no.="" 50-348,="" joseph="" m.="" farley="" nuclear="" plant,="" unit="" 1,="" houston="" county,="" alabama="" date="" of="" application="" for="" amendment:="" december="" 9,="" 1993,="" as="" supplemented="" february="" 23,="" and="" april="" 1,="" 1994="" brief="" description="" of="" amendment:="" the="" amendment="" modifies="" technical="" specification="" (ts)="" 3/4.4.6,="" steam="" generators,="" and="" ts="" 3/4.4.9,="" specific="" activity,="" and="" their="" associated="" bases.="" the="" steam="" generator="" plugging/="" repair="" limit="" is="" being="" modified="" in="" the="" ts="" to="" incorporate="" a="" 2.0="" volt="" steam="" generator="" tube="" support="" plate="" interim="" plugging="" criteria="" for="" cycle="" 13="" only.="" in="" addition,="" the="" ts="" limit="" for="" specific="" activity="" of="" dose="" equivalent="" i\131\="" and="" its="" transient="" dose="" equivalent="" i\131\="" reactor="" coolant="" specific="" activity="" will="" be="" reduced="" by="" a="" factor="" of="" 4="" in="" order="" to="" increase="" the="" allowable="" leakage="" in="" the="" event="" of="" a="" steam="" line="" break="" for="" cycle="" 13="" only.="" date="" of="" issuance:="" april="" 5,="" 1994.="" effective="" date:="" april="" 5,="" 1994.="" amendment="" no.:="" 106.="" facility="" operating="" license="" no.="" npf-2.="" amendment="" revises="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" january="" 19,="" 1994="" (59="" fr="" 2871)="" the="" february="" 23,="" 1994,="" and="" april="" 1,="" 1994,="" letters="" provided="" supplemental="" information="" and="" deleted="" the="" requested="" ts="" upper="" limit="" bobbin="" voltage="" of="" 5.7="" volts="" for="" tube="" plugging="" that="" was="" requested="" in="" the="" december="" 9,="" 1993,="" letter="" and="" retained="" the="" current="" value="" of="" 3.6="" volts.="" the="" february="" 23="" and="" april="" 1,="" 1994,="" supplements="" did="" not="" change="" the="" original="" no="" significant="" hazards="" consideration="" finding.="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 5,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" houston-love="" memorial="" library,="" 212="" w.="" burdeshaw="" street,="" post="" office="" box="" 1369,="" dothan,="" alabama="" 36302="" texas="" utilities="" electric="" company,="" docket="" nos.="" 50-445="" and="" 50-446,="" comanche="" peak="" steam="" electric="" station,="" units="" 1="" and="" 2,="" somervell="" county,="" texas="" date="" of="" amendment="" request:="" may="" 21,="" 1993,="" as="" supplemented="" by="" letter="" dated="" february="" 23,="" 1994.="" brief="" description="" of="" amendment:="" the="" amendments="" change="" the="" technical="" specifications="" by="" replacing="" the="" requirements="" associated="" with="" the="" control="" room="" heating="" and="" ventilation="" system="" with="" requirements="" related="" to="" operation="" of="" the="" control="" room="" filtration="" system="" and="" control="" room="" air="" conditioning="" system.="" the="" proposed="" change="" is="" consistent="" with="" the="" requirements="" of="" the="" westinghouse="" standard="" technical="" specifications="" (nureg-1431)="" issued="" on="" september="" 28,="" 1992.="" date="" of="" issuance:="" april="" 6,="" 1994="" effective="" date:="" april="" 6,="" 1994,="" to="" be="" implemented="" within="" 30="" days="" of="" issuance.="" amendment="" nos:="" unit="" 1--amendment="" no.="" 23;="" unit="" 2--amendment="" no.="" 9="" facility="" operating="" license="" nos.="" npf-87="" and="" npf-89:="" amendments="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" august="" 18,="" 1993="" (58="" fr="" 43933).="" the="" february="" 23,="" 1994,="" submittal="" provided="" supplemental="" information="" to="" the="" application="" and="" did="" not="" change="" the="" initial="" no="" significant="" hazards="" determination.="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 6,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" university="" of="" texas="" at="" arlington="" library,="" government="" publications/maps,="" 701="" south="" cooper,="" p.o.="" box="" 19497,="" arlington,="" texas="" 76019.="" union="" electric="" company,="" docket="" no.="" 50-483,="" callaway="" plant,="" unit="" 1,="" callaway="" county,="" missouri="" date="" of="" application="" for="" amendment:="" september="" 24,="" 1993.="" brief="" description="" of="" amendment:="" the="" amendment="" revises="" the="" technical="" specifications="" to="" extend="" the="" reporting="" period="" of="" the="" semiannual="" radioactive="" effluent="" release="" report="" from="" semiannually="" to="" annually.="" additionally,="" the="" report="" submission="" date="" changes="" from="" 60="" days="" after="" january="" 1="" and="" july="" 1="" of="" each="" year="" to="" before="" may="" 1="" of="" each="" year.="" the="" changes="" to="" the="" reporting="" period="" and="" report="" date="" are="" being="" made="" to="" implement="" the="" august="" 31,="" 1992,="" change="" to="" 10="" cfr="" 50.36a.="" the="" affected="" technical="" specification="" sections="" are="" 1.18,="" 3.11.1.4,="" 3.11.2.6,="" 6.9.1.7,="" 6.14c,="" and="" the="" index.="" date="" of="" issuance:="" april="" 14,="" 1994.="" effective="" date:="" april="" 14,="" 1994.="" amendment="" no.:="" 89.="" facility="" operating="" license="" no.="" npf-30.="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" march="" 2,="" 1994="" (59="" fr="" 10016).="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 14,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" callaway="" county="" public="" library,="" 710="" court="" street,="" fulton,="" missouri="" 65251.="" wisconsin="" public="" service="" corporation,="" docket="" no.="" 50-305,="" kewaunee="" nuclear="" power="" plant,="" kewaunee="" county,="" wisconsin="" date="" of="" application="" for="" amendment:="" february="" 1,="" 1994.="" brief="" description="" of="" amendment:="" the="" amendment="" revises="" the="" ts="" by="" removing="" the="" review="" of="" the="" emergency="" plan="" and="" its="" implementing="" procedures="" from="" the="" list="" of="" responsibilities="" of="" the="" plant="" operations="" review="" committee="" (porc).="" guidance="" for="" this="" change="" was="" provided="" in="" generic="" letter="" 93-07,="" ``modification="" of="" the="" technical="" specification="" administrative="" control="" requirements="" for="" emergency="" and="" security="" plans,''="" dated="" december="" 28,="" 1993.="" several="" other="" administrative="" ts="" changes="" were="" also="" made="" including="" removing="" specific="" titles="" from="" the="" list="" of="" porc="" members="" in="" ts="" 6.5.a.2="" and="" deleting="" ts="" 6.5.b="" which="" describes="" the="" corporate="" support="" staff.="" date="" of="" issuance:="" april="" 7,="" 1994.="" effective="" date:="" april="" 7,="" 1994.="" amendment="" no.:="" 107.="" facility="" operating="" license="" no.="" dpr-43.="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" march="" 2,="" 1994="" (59="" fr="" 10017)="" the="" commission's="" related="" evaluation="" of="" the="" amendment="" is="" contained="" in="" a="" safety="" evaluation="" dated="" april="" 7,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" university="" of="" wisconsin="" library="" learning="" center,="" 2420="" nicolet="" drive,="" green="" bay,="" wisconsin="" 54301.="" wisconsin="" public="" service="" corporation,="" docket="" no.="" 50-305,="" kewaunee="" nuclear="" power="" plant,="" kewaunee="" county,="" wisconsin.="" date="" of="" application="" for="" amendment:="" may="" 5,="" 1993="" as="" supplemented="" march="" 4,="" 1994.="" brief="" description="" of="" amendment:="" the="" amendment="" changes="" the="" kewaunee="" nuclear="" power="" plant="" (knpp)="" technical="" specifications="" (ts)="" in="" response="" to="" nrc="" generic="" letter="" 90-06.="" this="" letter="" deals="" with="" generic="" issue="" 70="" and="" generic="" issue="" 94,="" which="" focus="" on="" power-operated="" relief="" valve="" and="" block="" valve="" reliability="" and="" additional="" low-temperature="" overpressure="" protection.="" the="" amendment="" revises="" ts="" section="" 3.1="" by="" adding="" restrictions="" on="" the="" restart="" of="" an="" inactive="" reactor="" coolant="" pump,="" modifying="" the="" limiting="" conditions="" for="" operation="" of="" the="" pressurizer="" power-operated="" relief="" valves="" (porvs)="" and="" associated="" block="" valves,="" and="" adding="" provisions="" to="" ensure="" that="" adequate="" low-temperature="" overpressure="" protection="" (ltop)="" is="" available.="" additionally,="" this="" amendment="" modifies="" the="" limiting="" conditions="" for="" operation="" for="" reactor="" coolant="" temperature="" and="" pressure="" by="" adding="" figure="" ts="" 3.1-4="" to="" define="" 10="" cfr="" 50="" appendix="" g="" pressure="" and="" temperature="" limitations="" for="" ltop="" evaluation="" through="" the="" end="" of="" operating="" cycle="" 20.="" date="" of="" issuance:="" april="" 7,="" 1994.="" effective="" date:="" april="" 7,="" 1994.="" amendment="" no.:="" 108.="" facility="" operating="" license="" no.="" dpr-43.="" amendment="" revised="" the="" technical="" specifications.="" date="" of="" initial="" notice="" in="" federal="" register:="" july="" 21,="" 1993="" (58="" fr="" 39062)="" the="" march="" 4,="" 1994,="" submittal="" provided="" additional="" clarifying="" information="" and="" changed="" the="" ltop="" allowed="" outage="" time="" from="" 7="" days="" to="" a="" more="" conservative="" 5="" days.="" this="" modification="" 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="" april="" 7,="" 1994.="" no="" significant="" hazards="" consideration="" comments="" received:="" no.="" local="" public="" document="" room="" location:="" university="" of="" wisconsin="" library="" learning="" center,="" 2420="" nicolet="" drive,="" green="" bay,="" wisconsin="" 54301.="" dated="" at="" rockville,="" maryland,="" this="" 20th="" day="" of="" april="" 1994.="" for="" the="" nuclear="" regulatory="" commission.="" gus="" c.="" lainas,="" acting="" director,="" division="" of="" reactor="" projects--i/ii,="" office="" of="" nuclear="" reactor="" regulation.="" [fr="" doc.="" 94-10011="" filed="" 4-26-94;="" 8:45="" am]="" billing="" code="" 7590-01-p="">