[Federal Register Volume 59, Number 22 (Wednesday, February 2, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10202]
[[Page Unknown]]
[Federal Register: February 2, 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 January 7, 1994, through January 21, 1994.
The last biweekly notice was published on January 19, 1994 (59 FR
2859).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 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 March 4, 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.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: November 4, 1993
Description of amendment request: The proposed amendment revises
Technical Specification 6.13.1 to provide use of alarming dosimeters in
high radiation areas. This change includes newly revised 10 CFR Part 20
requirement references and is consistent with NUREG-1413, Standard
Technical Specifications - Westinghouse Plants, Specification 5.11.1.
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. This change could involve a reduction in
personnel radiation exposure by utilizing alarming dosimeters. This
change does not involve any plant systems or components which could
increase the probability of an accident. Therefore, there would be
no increase in 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. This change could reduce the possibility of an accidental
overexposure by alerting personnel when their maximum allowable
exposure has been received. This change does not involve any plant
systems or components. Therefore, the proposed changes do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety. Due to the nature of this
proposed change, it is not related to any plant system. 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: Hartsville Memorial Library,
Home and Fifth Avenues, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: S. Singh Bajwa
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 amendment request: August 27, 1993
Description of amendment request: The proposed amendments would
revise the Braidwood Station, Units 1 and 2, and Byron Station, Units 1
and 2, Technical Specifications (TS) regarding inspection requirements
for pipe snubbers. The proposed changes implement Generic Letter (GL)
90-09, and would affect the surveillance requirements of TS section
4.7.8 and the bases for these requirements, section 3/4.7.8.
Specifically, the amendment would change the existing inspection
periods, visual inspection acceptance criteria, and functional test
requirements. Additionally, there would be changes to the bases to
include reference to GL 90-09, and other editorial changes would be
made.
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 amended surveillance requirement adds a table that addresses
the maximum number of snubber failures that can be tolerated prior
to reducing the inspection interval. This number is a function of
the population size of a particular type of snubber. The revised
requirement will allow the inspection intervals to be compatible
with the 24 month fuel cycles, and provisions are included to extend
the inspection interval up to 48 months. A provision is included to
allow an evaluation to determine operability to justify continued
operation with a snubber that is unacceptable.
The purpose of the amendment request is to provide for
alternative inspection intervals that take the size of the
population of a snubber type into account. The proposed change
provides the same confidence level and allows snubber inspection and
corrective action to be performed during refueling outages. This
allows the plant to avoid a mid-cycle outage due to a small number
of snubber failures.
The proposed change allows for a small percentage of snubbers in
each category to fail the required visual examination without
adjusting the inspection frequency. If a statistically significant
percentage of snubbers fail, the visual examination inspection
interval is reduced based on the percentage of failed snubbers.
The proposed change has no direct or indirect impact on
reactivity management activities.
The change is not expected to have an impact on equipment
failures. Any snubbers that fail to meet the visual examination
acceptance criteria are either functionally tested in the as-found
condition to verify continued acceptability, or an evaluation is
performed to demonstrate the acceptability of continued operation
with an unacceptable snubber. No new equipment is being introduced
and no systems are operated in a configuration that has not been
evaluated, so no new failure modes are introduced.
The affected transients are the design basis earthquake and the
spectrum of event initiating transients, with the capability of
imposing significant dynamic loads or otherwise which impact the
structural integrity of the Reactor Coolant System (RCS).
The snubbers are installed to ensure the structural integrity of
the RCS and required support systems. Their failure is passive in
nature. The probability of a transient initiating event occurring is
unrelated to the existence or condition of equipment that is
designed to perform a mitigating function. The snubbers are
installed to ensure an acceptable system response to a dynamic load,
and their availability does not impact the frequency of occurrence
of earthquakes or other transients resulting in significant dynamic
loading.
The revised testing provisions are designed to allow some
flexibility while still maintaining a high probability that the
installed snubbers will be capable of performing their intended
function when required. The revised surveillances appropriately
consider the size of the population of a particular type of snubber,
and are sufficient to ensure the consequences of an accident will be
unchanged when the revised requirements are implemented. By
maintaining a statistically high level of confidence in the function
of the plant's snubbers, the system response to transient initiating
events will be as designed and thus, the off-site dose projected to
occur of any affected transient will remain acceptably low.
As previously stated, the revised surveillance provides a high
confidence that the affected systems will remain intact and
functional. Evaluation of the effects of operating with a degraded
snubber is required to ensure that adequate margin exists to support
continued plant operation. If this evaluation cannot adequately
justify continued operation, the appropriate action statement will
be applied. These provisions are sufficient to assure that the
probability of an equipment malfunction will not increase.
The consequences of equipment malfunction will not increase.
Sufficient redundancy exists to accommodate the complete failure of
one train of required equipment. The requisite electrical and
physical separation are sufficient to ensure that the redundant
train remains unaffected. This redundancy is adequate to ensure that
the undetected failure of a snubber will not have a severe impact on
overall system response to a transient.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The possibility of a new or different type of accident is not
created by this change. No new or different equipment is being
introduced, and no system will be operated in a different
configuration without first having the effects of the new
configuration evaluated. The new configuration would be system and/
or plant operation with a snubber installed that has failed its
visual examination. The required evaluation must be sufficient to
provide confidence that continued operation is acceptable;
otherwise, the provisions of the action statement will be observed.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
No reduction in the margin of safety will occur as a result of
this proposed change. As previously described, the controls in place
will provide a high confidence the affected systems will continue to
be functional. No significant increase in the rate of occurrence of
undetected inoperable snubbers is expected to occur, and the
allowable failures prior to applying an increased test frequency is
still a small percentage of the total snubber population.
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
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 amendment request: September 2, 1993, supplemented by
letter dated January 7, 1994.
Description of amendment request: The proposed amendment would
revise the Braidwood Station, Units 1 and 2, and Byron Station, Units 1
and 2, Technical Specifications (TS) to allow replacement of the 125
Volt DC Gould batteries with the new 125 Volt DC AT&T batteries and
rephrase their design duty cycle. In addition, the proposed amendment
would revise the batteries crosstie loading limitations and the
crosstie breaker limitations. The associated Bases would also be
revised to discuss the purpose for the crosstie limitations and to
discuss design duty cycle 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:
A. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The replacement AT&T battery has been selected to meet or exceed
the design, functional, and operational requirements of those of the
present Gould battery, including crosstie load limitations. The
crosstie breaker limitation change to allow crosstie between two
shutdown units is consistent with the Safety Evaluations issued with
Technical Specification Amendment 5 for both Braidwood and Byron
Stations. The remaining changes are administrative in nature or
provide clarification to maintain consistency with other Technical
Specifications and the Standard Technical Specifications.
The overall design, function, and operation of the DC system and
equipment has not been altered by these changes. The proposed
changes do not affect any accident initiators or precursors and do
not alter the design assumptions for the systems or components used
to mitigate the consequences of an accident as analyzed in UFSAR
Chapter 15. Therefore, there is no increase in the probability or
consequences of an accident previously evaluated.
B. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The replacement AT&T battery will provide the same functions as
those of the present Gould battery and will be operated with the
same types of operational controls. These limits will include
battery float terminal voltage, individual cell voltage and
electrolyte specific gravity, and crosstie loading. Crosstie
conditions are allowed under the present Technical Specifications.
The remaining changes are administrative in nature or provide
clarification to maintain consistency with other Technical
Specifications and the Standard Technical Specifications.
The DC system and its equipment will continue to perform the
same functions and be operated in the same fashion. The proposed
change does not create any new or common failure modes. The proposed
changes do not introduce any new accident initiators or precursors,
or any new design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, the possibility
of a new or different kind of accident from any accident previously
evaluated has not been created.
C. The proposed change does not involve a significant reduction
in a margin of safety.
The replacement AT&T battery will meet or exceed the design,
functional, and qualification requirements of those of the present
Gould batteries. The proposed Technical Specification limitations
for the AT&T battery are derived from the same methodology and
margins as those for the Gould battery. Increasing the crosstie
loading limit takes advantage of the larger AT&T battery capacity
with its increased design margin. The proposed change to the
crosstie loading limit will continue to conservatively envelop the
postulated design requirements. The remaining changes are
administrative in nature or provide clarification to maintain
consistency with other Technical Specifications and the Standard
Technical Specifications.
The inherent design conservatism of the DC system and its
equipment has not been altered. The DC system and its equipment will
continue to be operated with the same degree of conservatism.
Therefore, there is no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: October 28, 1993
Description of amendment request: The proposed amendment would
revise the Emergency Core Cooling System (ECCS) injection valve stroke
times and ECCS response times for Motor-Operated Valve (MOV)
modifications that increase injection valve stroke times.
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
because:
The probability of an accident previously evaluated will not
increase as a result of this change, because the only modification
being performed is to the stroke times for the LPCS [Low Pressure
Core Spray System], LPCI [Low Pressure Coolant Injection System],
and HPCS [High Pressure Core Spray System] injection valves.
Changing the opening or closing time of the injection valves for
these ECCS systems does not cause any accident previously evaluated
to occur. Therefore, modifying their stroke times will not increase
the probability of occurrence for any accident previously evaluated.
The consequences of a LOCA [Loss of Coolant Accident] are not
significantly increased and do not exceed the previously accepted
licensing criteria for this accident. GE [General Electric Company]
has calculated the revised licensing basis PCT [Peak Centerline
Temperature] for LaSalle Station to be 1260 deg.F, which is well
below the 2200 deg.F criterion of 10 CFR 50.46 and Section 15.6.5 of
NUREG-0800 (Standard Review Plan). The acceptance criteria for
cladding oxidation, metal-water reaction (hydrogen generation),
coolable geometry and long-term cooling also continue to be met with
the increased valve stroke times.
GE has performed sensitivity analysis justifying the continued
applicability of previous analyses for Anticipated Transients
Without Scram (ATWS), containment analyses, off-site dose (Main
Steamline Break Outside Containment), and HPCS-related transients
(Loss of Feedwater Flow). Other events are not affected because
these systems are not assumed to function.
2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because:
The only modification is to increase the stroke time of the
injection valves for LPCS, LPCI, and HPCS. This does not result in
any changed component interactions, other than to increase the
affected ECCS response times. The injection valves will still
provide the function for which they were designed. Since the systems
will continue to function as intended, the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3) The proposed changes do not involve a significant reduction
in a margin of safety because:
While the calculated licensing basis PCT is larger than that
previously calculated with the current valve stroke times, the new
PCT remains far below the 2200 deg.F licensing acceptance limit
associated with a LOCA. This limit has been previously evaluated as
providing a sufficient margin of safety. All other LOCA licensing
limits also continue to be met with the increased stroke times. For
other accidents and transients, the increased stroke times have a
negligible effect on the results, so the margin of safety is
preserved.
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: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: James E. Dyer
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: April 15, 1992, as modified by letters
dated December 8, 1992, and June 25, 1993.
Description of amendment request: The proposed amendment would
revise the provisions in the Technical Specifications to incorporate
Generic Letter 90-06, ``Resolution of Generic Issue 70, `Power-Operated
Relief Valve and Block Valve Reliability,' and Generic Issue 94,
`Additional Low-Temperature Overpressure Protection for Light-Water
Reactors,''' power-operated relief valve (PORV) requirements for power
operation, and to modify the primary coolant system (PCS) overpressure
protection specification venting 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:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed Technical Specification requiring Operability of
the PORVs and their block valves does not alter plant operation or
configuration in any way. It is current practice to maintain these
valves in an Operable condition to meet the requirements of existing
Specification 3.1.8, which is applicable when below 430 deg.F. The
effect of the proposed changes is to extend the applicability of the
Operability requirement for these valves. The addition of PORV
Operability requirements when at Hot Standby and above would not
involve a significant increase in the probability or consequence of
an accident previously evaluated.
Replacing the requirements to vent the PCS through a 1.3 square
inch vent with a vent capable of relieving 167 gpm at a pressure
less than the Appendix G limits will not significantly increase the
probability or consequences of an overpressurization event
occurring. The 1.3 square inch vent area in Technical Specification
3.1.8 was intended to be a means of protecting the Primary Coolant
System (PCS) from exceeding the limit of the 10 CFR [Part] 50,
Appendix G, curve following an overpressure transient. Analysis has
shown that manual vent valves PC-514 and PC-515 will provide a
relief capacity of 167 gpm at a PCS pressure of approximately 115
psig, well below the minimum 331 psig limit (Appendix G curve limit
for a 40 deg.F/hr heat-up). This relief capacity will protect the
PCS against a pressure transient caused by a maximum charging/
letdown imbalance coincident with a 40 deg.F/hr PCS heat-up rate and
a 60 deg.F/hr pressurizer heat-up rate.
Two other pressure transients, a High Pressure Safety Injection
(HPSI) pump start and a Primary Coolant Pump (PCP) start, are also
precluded. With the PCS in a vented and depressurized state, the PCS
would be below 212 deg.F Existing technical specifications require
both HPSI pumps to be rendered inoperable below 260 deg.F and, with
the system depressurized, normal operating procedures prohibit a PCP
start due to insufficient pump net positive suction head (NPSH).
Therefore, the 1.3 square inch requirement can be replaced with
a requirement to have a vent capable of relieving 167 gpm at a PCS
pressure less than or equal to the Appendix G limit with no
significant increase in the probability or consequences of an
overpressurization event occurring.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The addition of PORV Operability requirements when at Hot
Standby and above will not alter plant operation or configuration.
It will not alter any equipment or analyses. Therefore the addition
of these PORV Operability requirements will not create the
possibility of a new or different kind of accident from any
previously evaluated.
The new technical specifications requirements for PCS vent
capacity will provide an equivalent or better, overpressure
protection as compared to the existing requirement. No analysis has
been found that shows that the existing 1.3 square inch vent area
will protect the PCS from exceeding the Appendix G curve limit.
However, analysis has been developed which shows that manual vent
valves PC-514 and PC-515 will provide adequate relief capacity,
maintaining PCS pressure within the 10 CFR [Part] 50, Appendix G,
limits. Furthermore, two other pressure transients, a HPSI pump
start and a PCP start, are also precluded by either existing
technical specifications or normal plant operating procedures.
Another related analysis has shown that relief valve RV-3164,
the Low Pressure Safety Injection (LPSI) pumps, the LPSI pump seals,
and the system piping of the shutdown cooling system have the
capability of providing adequate overpressure protection to the
shutdown cooling system.
The addition of the manual vent valves do not introduce a vent
path where a vent path had not previously existed. Therefore, the
possibility of an accident of a new or different type, than
previously evaluated in the FSAR, will not be created.
3. Involve a significant reduction in a margin of safety.
The margin of safety will not be reduced by the proposed
Technical Specifications changes. The extension of PORV Operability
requirements has no effect on any margin of safety. The previous
requirement assumed that a vent with an equivalent flow area as the
original PORV would provide the same relief as the PORV itself and
gave no consideration to how that flow area should factor in system
losses or vent location. The new technical specification requirement
offers a means to ensure the PCS will be protected against all
achievable overpressure transients for the system configuration,
with analyses to support it.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: A. Randolph Blough, Acting
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: December 16, 1993
Description of amendment request: The proposed amendment clarifies
the requirements for maintaining secondary containment integrity when
one or more Reactor Building Ventilation supply and exhaust valves are
declared inoperable. The technical specification (TS) change adds a new
Limiting Condition for Operation, Basis Statement, and Surveillance
Requirements for these isolation valves. The change revises TS
definition 1.14c, adds new Specifications 3.5.B.2, 3.5.B.3, 4.5.R, and
a Basis statement to TS 3.5, edits T.S. 3.5.B.1.1. It also renumbers TS
3.5.B.2 through 3.5.B.4, to 3.5.B.4 through 3.5.B.7. It also revises
specification references within to reflect new specification numbers.
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:
GPU Nuclear has determined that operation of the Oyster Creek
Nuclear Generating Station in accordance with the proposed Technical
Specifications does not involve a significant hazard. The changes do
not:
1. Involve a significant increase in the probability or the
consequence of an accident previously evaluated.
The failure of any component in the [Reactor Building
Ventilation System] RBVS was not considered as a credible initiating
event for a design basis accident. However, the RBVS is designed to
mitigate the consequences of a potential radiological release by the
isolation of all supply and exhaust ducts to the environs. Since the
failure of the RBVS was never considered as one of the initiators of
an accident, this proposed change cannot increase the probability of
occurrence of an accident. During the proposed Limiting Condition
for Operation (LCO), the supply or exhaust duct will be isolated
within 8 hours by one isolation valve secured in its post accident
design position. Since the duct can perform its post accident design
function (isolation), there is no increase in the consequences of an
accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The design function of the RBVS automatic isolation valves is to
isolate the ducts which penetrate the Reactor Building or Secondary
Containment during a radiological release. During the LCO, the duct
will be isolated within 8 hours by one isolation valve secured in
the closed position. Since the duct will be isolated, this change
will not create a possibility for an accident or malfunction of a
different type than previously identified.
3. Involve a significant reduction in a margin of safety.
If a RBVS automatic isolation valve (supply) is declared
inoperable, the proposed LCO would allow continued plant operation
with that supply duct isolated. Since the RBVS can still perform its
design function (redundant ductwork) under normal plant and design
accident conditions, there is no reduction in the margin of safety.
For an inoperable isolation valve in the exhaust duct, the exhaust
duct will be isolated within 8 hours by one isolation valve secured
in the closed position. Further, the RBVS and the [Standby Gas
Treatment System] SGTS will be aligned for an accident condition
with no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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 Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: November 17, 1993
Description of amendment requests: The proposed amendments would
modify the Technical Specifications to allow a portion of the Waste Gas
Holdup System (WGHS) Explosive Monitoring System to be inoperable for
160 days on a one-time basis. This is to allow replacement of the Waste
Gas Oxygen Analyzer.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
We have evaluated the proposed T/S changes and have determined
that the changes should involve no significant hazards
consideration. Operation of the Cook Nuclear Plant in accordance
with the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The purpose of the hydrogen and oxygen monitors is to
measure the concentrations of these gases in the WGHS to ensure that
the gas mixture is non-flammable. We can accomplish this purpose and
ensure safe operation of the WGHS by operating the system in the
proposed manner. During the equipment replacement interval, we will
be assuming that the hydrogen concentration is above the flammable
limit (4%). The limiting factor then is the oxygen concentration at
which hydrogen and oxygen become flammable. The existing hydrogen
monitor will be in continuous operation to verify the hydrogen
level. The information provided by the oxygen monitor being replaced
is not essential to the safe operation of the WGHS since it is
redundant to information provided by the remaining oxygen monitor.
The only difference between the two (2) oxygen monitors is that the
one being replaced provides an automatic isolation of the waste gas
decay tank when the oxygen concentration reaches 3%. The isolation
of the waste gas decay tanks will be performed manually during the
replacement. In the event the remaining oxygen monitor becomes
inoperable, we will follow the currently approved T/Ss. Since
operation of the WGHS in the manner we have proposed will ensure
that the purpose of the oxygen and hydrogen monitors is fulfilled
and safe operation of the WGHS is maintained, the proposed change
will not involve a significant increase in the probability or
consequences of a previously analyzed incident.
The proposed change to the Automatic Gas Analyzer (QC-31) tag
number to QC-1400 will not reduce in any way requirements or
commitments in the existing T/Ss. The proposed change will eliminate
confusion of spare parts of the new analyzer panel installed in
1990.
(2) Create the possibility of a new or different kind of
accident from any previously analyzed.
The proposed amendment does not create the possibility of a new
or different kind of accident from any previously evaluated. During
the replacement of the monitors the waste gas holdup system will
continue to operate normally. The proposed method of operation will
ensure that the oxygen and hydrogen gas mixture is non-flammable.
For this reason, operating the explosive gas monitoring system in
the proposed manner will not place the plant in a new or unanalyzed
condition. Therefore, we believe that this change will not introduce
a new or different kind of accident than previously analyzed.
The proposed editorial change will not create the possibility of
a new or different kind of accident from any previously evaluated,
because these changes will not place the plant in a new or
unanalyzed condition.
(3) Involve a significant reduction in a margin of safety.
The proposed amendment does not involve a significant reduction
in the margin of safety. The remaining oxygen monitor will be
available to maintain the oxygen concentration below the limit
required for hydrogen flammability in oxygen. In addition, the
oxygen grab samples will provide redundant information and will
serve as a check of the monitor's readings. If the remaining monitor
becomes inoperable, we will follow the actions of our current T/Ss.
During the equipment replacement period, we will be assuming that
the hydrogen concentration is above the flammable limit (4%). This
will then make the oxygen level the controlling parameter in a
possible flammable combination of oxygen and hydrogen. The existing
hydrogen monitor will be in continuous operation to verify the
hydrogen level. These proposed interim measures will not
significantly affect our ability to maintain the hydrogen and oxygen
concentration within the limits to prevent flammability. Therefore,
we believe that operation of the system in this manner does not
involve a significant reduction in a margin of safety.
The proposed editorial change will not involve a significant
reduction in margin of safety, because all accident analyses and
nuclear design bases remain unchanged.
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 requests involve 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: A. Randolph Blough, Acting
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: December 20, 1993
Description of amendment requests: The proposed amendments would
revise the Technical Specifications to change Train A and B Emergency
Loads from 8 hour to composite 4 hour, delete a load on the Train B
batteries load list, and revise the operational loads on the Train N
batteries.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed amendment 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.
(2) create the possibility of a new or different kind of
accident from any previously analyzed.
(3) involve a significant reduction in a margin of safety.
Criterion 1
The change is proposed to comply with the LOCA/LOOP [loss-of-
coolant accident/loss-of-offsite power] and SBO [Station Blackout]
requirement for Cook Nuclear Plant for battery testing. The
composite test as addressed above meets these requirements for four
hour test profiles. This change is consistent with the UFSAR three
hour LOCA/LOOP and NUMARC 87-00 Station Blackout Rule four hour.
Based on these considerations, the proposed change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2
The change only addresses the battery profile test which meets
both LOCA/LOOP and SBO for Cook Nuclear Plant. No specific physical
or operational changes to the plant will occur due to this change.
Thus, the change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Criterion 3
The proposed change revises the battery profile test table from
an eight hour to a four composite test which complies with both
LOCA/LOOP and SBO as defined for Cook Nuclear Plant. 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 requests involve 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: A. Randolph Blough, Acting
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: December 22, 1993
Description of amendment requests: The proposed amendments would
revise the Technical Specifications section addressing steam generator
stop valves making it more consistent with the revised Standard
Technical Specifications and clarifying certain 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:
Per 10 CFR 50.92, 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,
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.
Criterion 1
The limiting conditions for operation involving the steam
generator stop valves are not altered by this proposed change. The
surveillance requirements are lessened for Unit 2 in that valve
stroke timing does not have to be performed on valves that are
closed. This is consistent with the wording of the Unit 1 T/S, and
reflects the fact that, when closed, the valves are already in the
position required by the assumptions in the safety analysis and
therefore stroke timing is not necessary. The remaining changes are
consistent with NUREG 1431, and as such, 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 of the steam
generator stop valves are proposed as part of this amendment
request. The proposed changes do not involve any physical changes to
the plant. The changes will allow operation in Modes 2 and 3 with
more than one steam generator stop valve inoperable. However,
inoperable valves must be closed and their closure periodically
reverified. When closed, the valves are already in the position
required by the assumptions in the safety analysis. Thus, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3
The limiting conditions for operation involving the steam
generator stop valves are not altered by this proposed change. The
surveillance requirements are lessened for Unit 2 in that valve
stroke timing does not have to be performed on valves that are
closed. This is consistent with the wording of the Unit 1 T/S, and
reflects the fact that, when closed, the valves are already in the
position required by the assumptions in the safety analysis and
therefore stroke timing is not necessary. The remaining changes are
consistent with NUREG 1431, and, as such, have already been found
acceptable by the NRC. Therefore, it is concluded that 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 requests involve 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: A. Randolph Blough, Acting
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: December 22, 1993
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.4.e (Emergency Ventilation
System). TS 3.4.4.e currently permits fuel handling operations to
continue during refueling for up to 7 days with one circuit of the
emergency ventilation system inoperable provided all active components
of the other emergency ventilation system circuit are operable. The
proposed revision would permit fuel handling operations to continue
during refueling beyond 7 days with one circuit of the emergency
ventilation system inoperable provided the remaining emergency
ventilation system circuit is operable and in operation. The licensee
stated that the proposed revision is consistent with recently issued
Amendment No. 47 to the Nine Mile Point Unit 2 TSs and with the NRC's
Improved Standard Technical Specifications, NUREG-1433.
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 operation of Nine Mile Point Unit 1 [NMP1], in accordance
with the proposed amendment, will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Emergency Ventilation System responds to a release of
radioactivity to the secondary containment by maintaining a negative
pressure in secondary containment and by providing a filtered
elevated release. The proposed change to LCO [Limiting Condition for
Operation] 3.4.4.e would allow continuation of refueling beyond
seven days with one emergency ventilation circuit inoperable
provided the operable emergency ventilation circuit is in operation.
A plant specific PRA [Probabilistic Risk Assessment] was performed
to evaluate the probability of a fuel bundle drop event resulting in
a need to start the Emergency Ventilation System with a concurrent
failure of the Emergency Ventilation System that would result in an
unfiltered ground level release under the current Technical
Specifications and the proposed change. The results of this
assessment indicate that the probability is not significantly
increased. In addition, the order of magnitude of the probability of
such a release, under the current or proposed Technical
Specifications, is very small, i.e., 10-6. This amendment
requires no physical changes to NMP1. Therefore, the proposed
changes to the Technical Specifications do not significantly
increase the probability of an accident previously evaluated.
Section XV.C.3 of the UFSAR [Updated Final Safety Analysis
Report] evaluates a fuel bundle drop accident. The radiological
consequences of this accident are within the guidelines of 10 CFR
Part 100. The UFSAR radiological evaluation takes credit for the
operation of an emergency ventilation circuit in mitigating the
consequences of this accident. During refueling with one emergency
ventilation circuit inoperable for more than seven days, the
proposed Technical Specification change would require that an
operable emergency ventilation circuit be placed in operation. With
an operable emergency ventilation circuit operating prior to a fuel
bundle drop accident, the radiological consequences of this accident
remains bounded by the current UFSAR evaluation. Therefore, from a
radiological perspective, the proposed Technical Specification
change is bounded by the current radiological evaluation in the
UFSAR. Therefore, the Technical Specification change does not
significantly increase the consequences of a previously evaluated
accident.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This amendment does not involve any accident precursors or
initiators. During an accident involving the release of
radioactivity to the secondary containment atmosphere an operable
emergency ventilation circuit would already be running and
performing its safety function. The operating status of a running
emergency ventilation circuit, which was manually started, would be
unaffected by the receipt of an automatic start signal due to the
detection of high radiation in secondary containment. Accordingly,
the proposed Technical Specification change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with
proposed amendment, will not involve a significant reduction in a
margin of safety.
The current Technical Specifications, LCO 3.4.4, provide a
margin of safety by requiring both emergency ventilation circuits to
be operable during a refueling condition. With one emergency
ventilation circuit inoperable, the current Technical Specifications
allow continuation of refueling for up to seven days, at which time
refueling must be stopped. This Technical Specification requirement
ensures that an emergency ventilation circuit will be available to
provide a filtered release to the environment during an accident
which could result in the release of radioactivity to the secondary
containment atmosphere.
The proposed change to LCO 3.4.4.e would allow continuation of
refueling beyond seven days with one emergency ventilation circuit
inoperable provided the operable emergency ventilation circuit is in
operation. By placing the remaining operable emergency ventilation
circuit in operation, active single failures associated with its
startup have been eliminated. These eliminated failures include
automatic initiation instrumentation, relaying logic, breaker
operation, fan startup and valve operation. With an operable
emergency ventilation circuit in operation, its safety function is
being performed. In addition, the status of the operating emergency
ventilation circuit is indicated in the control room. Therefore, the
running, operable emergency ventilation circuit provides a level of
safety comparable to two non-running, operable emergency ventilation
circuits.
Based upon the above analysis, the margin of safety is not
significantly reduced by allowing refueling to continue beyond seven
days with one emergency ventilation circuit inoperable since the
operable emergency ventilation circuit is in operation.
These changes are consistent with Amendment No. 47 for Nine Mile
Point Unit 2 and with the Improved Standard Technical
Specifications, NUREG-1433.
Accordingly, as determined by the analysis above, this proposed
amendment involves no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Robert A. Capra
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: December 27, 1993
Description of amendment request: The proposed amendment would
relocate Technical Specification (TS) Tables 3.2.7, ``Reactor Coolant
Isolation Valves,'' and 3.3.4, ``Primary Containment Isolation
Valves,'' from TS 3.2.7/4.2.7 and 3.3.4/4.3.4, respectively, to a plant
procedure overning lists removed from TSs per Generic Letter (GL) 91-
08, ``Removal of Component Lists from Technical Specifications.'' The
plant procedure would be subject to the requirements specified in the
Administrative Controls section of the Nine Mile Point Nuclear Station
Unit No. 1 (NMP-1) TS. The proposed amendment would also make
conforming changes to the TS Bases. These lists of valves will continue
to be included in the NMP-1 Updated Final Safety Analysis Report
(FSAR). The licensee stated that the proposed changes would be
consistent with NRC staff guidance issued in GL 91-08.
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 operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
1.The proposed amendment does not involve a physical change to
any system, structure, or component that affects the probability or
consequences of any accident or malfunction of equipment important
to safety.
Relocation of the component lists to plant procedures and the
Updated FSAR is in accordance with Generic Letter 91-08. This change
does not alter the existing operability or surveillance requirements
for the components to which they apply. The proposed changes are
under the change control provisions in Section 6.0, ``Administrative
Controls,'' of the Technical Specifications. The changes associated
with the Bases for Specification[s] 3.2.7/4.2.7 and 3.3.4/4.3.4 are
consistent with the issuance of prior license amendments. Since the
proposed amendment does not affect the operation or testing of any
plant systems or components, it will have no impact on the
probability or consequences of accidents or malfunctions previously
evaluated.
2.The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes to Technical Specification 3.2.7/4.2.7,
3.3.4/4.3.4 and Bases do not introduce any new modes of plant
operation or new accident precursors, involve any physical
alterations to plant configurations, or make changes to system
setpoints which could initiate a new or different kind of accident.
The proposed changes relocate Reactor Coolant Primary Containment
Isolation Valve Tables 3.2.7 and 3.3.4 into a procedure governing
controlled lists removed from TS per Generic Letter 91-08 under the
change control provisions in Section 6.0, ``Administrative
Controls,'' of the Technical Specifications. The testing associated
with these valves remains unchanged, therefore, it will not affect
system or component operability. In addition, the removal of generic
reference to the 60 second closure time is consistent with
previously issued license amendments and has no impact on either the
Limiting Condition for Operation or Surveillance Requirement.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
The Technical Specification Limiting Conditions for Operation
and Surveillance Requirements for the valves listed in Tables 3.2.7
and 3.3.4 are not being altered. The valve lists will be
incorporated into a procedure governing controlled lists removed
from TS per Generic Letter 91-08. This is controlled by Section 6.0,
``Administrative Procedures.''
In addition, removal of generic reference to the 60 second
closure time is consistent with previously issued license amendments
and has no impact on either the Limiting Conditions for Operation or
Surveillance Requirements. Therefore, the proposed changes will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Robert A. Capra
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: October 4, 1993
Description of amendment request: The proposed amendment would
delete the emergency diesel-generator engine speed specification from
Surveillance Requirement (SR) 4.8.1.1.2a.5 and would replace the diesel
engine speed requirement with an electrical frequency requirement in SR
4.8.1.1.2g. Currently, SR 4.8.1.1.2a.5 specifies both a minimum engine
speed and a nominal electrical frequency and acceptable deviation from
the nominal value. SR 4.8.1.1.2g currently specifies only a minimum
engine speed. The specified minimum engine speed is not consistent with
the acceptable frequency deviation below the nominal value.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. 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.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10
CFR 50.92(c)(1)) because the proposed changes merely delete the
specification of emergency diesel-generator minimum engine speed
while retaining or substituting the specification of electrical
frequency to be attained by the emergency diesel-generator. The
diesel engine and generator are directly coupled and must rotate at
the same speed, therefore, speed and frequency are directly related
and specification of acceptance values for both parameters is
redundant. Furthermore, electrical frequency, not engine speed, is
the parameter of interest with regard the ability of the system to
power emergency loads. The proposed changes do not affect the
adequacy of the surveillance test or the reliability of the system
to power emergency loads, and do not involve any physical changes to
facility structures, systems, or components. Therefore, since the
reliability of the emergency diesel-generators will not be reduced,
the probability or consequences of any accident previously evaluated
is not increased.
B. The changes do not create the possibility of a new or
different kind
of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because no physical changes to facility structures,
systems, or components are involved and they do not affect the
manner by which the facility is operated.
C. The changes do not involve a significant reduction in a
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes
do not affect the manner by which the facility is operated or
involve changes to equipment or features which affect the
operational characteristics of the facility. 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: Exeter Public Library, 47
Front Street, Exeter, New Hampshire 03833.
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston Massachusetts 02110-2624.
NRC Project Director: John F. Stolz
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 proposed amendment would
implement 13 of the 47 line item Technical Specification (TS)
improvements recommended by Generic Letter 93-05. Most of the proposed
changes would revise the allowable time intervals for performing
certain Surveillance Requirements (SR) on various plant components
during power operation or would delete the requirement entirely or
under certain conditions. One proposed change would modify testing
requirements identified in an ACTION statement. The specific changes
are as follows:
1. SR 4.1.3.1.2 - The allowable interval between tests to
demonstrate the operability of any partially or fully withdrawn control
rod would be increased to 92 days from 31 days.
2. SR 4.6.4.1 - The allowable interval between tests to demonstrate
the operability of the hydrogen monitors by performing an Analog
Channel Operational Test (ACOT) would be increased to 92 days from 31
days, and by performing a Channel Calibration to every refueling outage
from 92 days on a staggered basis.
3. SR 4.3.2.1, Table 4.3-2, Functional Unit 3.c.4 and SR 4.3.3.1,
Table 4.3-3, Functional Units 1 through 6 - The allowable interval
between tests to demonstrate the operability of the radiation monitors
by performing an ACOT and Digital Channel Operational Test (DCOT) would
be increased to quarterly from monthly.
4. SR 4.4.6.2.2- The time the plant may be in Cold Shutdown before
Pressure Isolation Valve testing is required prior to entering Mode 2
would be increased to 7 days from 72 hours.
5. SR 4.4.11.1 - The allowable interval between tests to
demonstrate the operability of the Reactor Coolant System vent block
valves would be increased to cold shutdown from every 92 days.
6. SR 4.4.3.2 - The allowable interval between tests to verify
pressurizer heater capacity would be increased to each refueling outage
from 92 days.
7. SR 4.5.1.1.1 - The requirement to verify the boron
concentration of the accumulator contents after a volume increase of
1-percent or more would be removed under certain conditions.
8. SR 4.5.1.1.2 - The requirement to perform an ACOT and Channel
Calibration on accumulator water level and pressure instrumentation
would be deleted.
9. SR 4.5.2 - The requirement to visually inspect the containment
sump upon completion of each containment entry would be modified to
avoid unnecessary containment sump inspections when multiple
containment entries are made on the same day.
10. SR 4.6.2.1 - The allowable interval between tests to verify
that each containment spray nozzle is unobstructed would be increased
to every 10 years from 5 years.
11. SR 4.6.4.2 - The allowable interval between tests to
demonstrate operability of each hydrogen recombiner system would be
increased to each refueling interval from 6 months.
12. SR 4.7.1.2.1 - The allowable interval between tests of the
auxiliary and startup feedwater pumps would be increased to 92 days on
a staggered test basis from 31 days.
13. TS 3.8.1.1 - The ACTION statements would be changed so that
when the Limiting Conditions for Operation are not met due to:
a. Inoperability of one or two offsite power circuits the starting
of a diesel-generator would no longer be required;
b. Inoperability of one diesel-generator, starting of the remaining
diesel-generator would be required within 8 hours only under certain
conditions instead of within 24 hours under all conditions;
c. Inoperability of a diesel-generator and an offsite power
circuit, starting of the remaining diesel-generator would be required
within 8 hours only under certain conditions instead of within 24 hours
under all conditions.
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.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)) because the proposed changes either merely modify the
allowable intervals between certain surveillance tests, delete the SR,
or alter an ACTION statement with regard to required testing. The
safety functions of the related structures, systems, or components are
not changed in any manner nor are the reliabilities of any structure,
system, or component reduced by the revised surveillance or testing
requirements. The changes do not affect the manner by which the
facility is operated and do not change any facility design feature,
structure, system, or component. Since there is no change to the
facility or operating procedures, and the safety functions and
reliabilities of structures, systems, or components are not affected,
there is no affect upon the probability or consequences of any accident
previously analyzed.
B. The changes do not create the possibility of a new or different
kindof accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because they do not change the facility or affect the
manner by which the facility is operated. The proposed changes merely
change certain surveillance or testing requirements.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because they do not affect the manner by
which the facility is operated or change equipment or features which
affect the operational characteristics of the facility.
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: Exeter Public Library, 47
Front Street, Exeter, New Hampshire 03833.
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston Massachusetts 02110-2624.
NRC Project Director: John F. Stolz
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: December 17, 1993
Description of amendment request: The proposed amendment changes
the action statements for the limiting conditions associated with the
electrical power sources (Technical Specification 3.8.1.1).
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:
In accordance with 10 CFR 50.92, NNECO has reviewed the attached
proposed changes and has concluded that they do not involve a
significant hazards consideration. The basis for this conclusion is
that the three criteria of 10CFR50.92(c) are not compromised. The
proposed changes do not involve a significant hazards consideration
because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to rewrite the action statements for
Technical Specification 3.8.1.1 will decrease the wear on the EDGs
[emergency diesel generators] by reducing the number of required
starts. These changes will also allow adequate time for the
completion of manufacturer recommended EDG engine prelube and warmup
procedures. They ensure that the operability of the AC sources is
demonstrated with reasonable assurance. Also, the reliability of the
EDGs will be enhanced by reducing the potential for numerous
unloaded EDG starts during an LCO [limiting condition for operation]
period. The proposed changes could reduce the number of required
unloaded EDG starts from nine to one. Therefore, these proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to rewrite the action statements for
Technical Specification 3.8.1.1 do not alter the method of operating
the plant, nor do they introduce any new failure modes. The proposed
changes affect EDG testing frequency only, they have no impact on
any accident analysis. The proposed changes provide assurance that
the EDGs will be able to power their respective safety systems if
required. Also, they do not involve any physical alterations to
plant equipment or procedures which would introduce any new or
unique operational modes or accident precursors.
3.Involve a significant reduction in a margin of safety.
The proposed changes to rewrite the action statements for
Technical Specification 3.8.1.1 do not affect the capability of the
EDGs to perform their function. The intent of the changes is to
increase the overall EDG reliability, by reducing the wear resulting
from excessive and unwarranted testing. 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: 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
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: June 12, 1992, as supplemented September
17, 1992, March 17, 1993, August 17, 1993, August 18, 1993, and
December 29, 1993.
Description of amendment request: The purpose of the proposed
changes is to revise the Technical Specifications (TS) to permit
operation of the James A. FitzPatrick Nuclear Power Plant at an uprated
power of 2536 MWt. The licensee's engineering analyses and evaluations
confirm that the plant can be operated at an uprated power. The
increase in the rated power from 2436 MWt to 2536 MWt corresponds to a
4.8 percent increase in rated steam flow. The increase in rated power
remains below the plant design power level of 2550 MWt which was the
basis for the original plant safety evaluation.
The changes affect the operating parameters of the reactor,
operational restrictions, setpoints for safety systems, analytical
results, and test requirements. There are also administrative changes.
The changes in each of these categories are summarized as follows:
Reactor Parameters: The effect on reactor parameters is limited.
Higher power is achieved by control rod pattern adjustments to increase
reactor thermal power in a more uniform (flattened) powerdistribution
to increase steam flow without increasing core recirculation flow. This
requires an increased reactor dome pressure for adequate turbine inlet
pressure.
Operational Limits: The increased thermal power requires a change
to the limitation on operation in the high power low flow portion of
the power/flow map to limit thermal hydraulic instabilities and power
oscillations.
Setpoints: The increased reactor pressure has a direct impact on
the high pressure scram setpoint and the safety relief valve setpoint.
Additionally, the bypass for the turbine stop valve closure and control
valve fast closure scram will be changed in proportion to the increase
in thermal power.
Analysis Results: Analyses of uprated power transients and
accidents requires changes to various TS and their Bases. Operational
parameters and assumptions used in analyses were revised to reflect
their use as initial conditions. Revised radiological analyses changed
dose results. The results of the accident analyses requires revisions
to properly reflect plant capabilities.
Testing: A number of changes to testing requirements result from
power uprate. The increase to reactor pressure has a direct effect on
hydrostatic leakage testing pressure. The test pressure for High
Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling
(RCIC) pumps will be revised to reflect safety relief valve setpoints
assumed in analyses.
Administrative: Administrative changes (i.e., adding references,
revising references and correcting associated errors) will also be
made.
No changes to the Radiological Effluent Technical Specifications
were identified.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant at a thermal power of 2536
MWt will not involve a significant hazards consideration as defined
in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The James A. FitzPatrick nuclear power plant was reviewed for
operation at a rated power of 2550 MWt at the time of its operating
license, [***]. This review was based on the original design of the
plant. Since that time, a number of safety issues of a generic and
plant specific nature as well as plant modifications have changed
the originally reviewed design.
Generic criteria, methodologies and evaluation scope required to
uprate BWRs up to 5% were prepared by General Electric and submitted
to the NRC in LTR-1 [NEDC-31897P-A, ``Generic Guidelines for General
Electric Boiling Water Reactor Power Uprate]. This was
supplemented by the submittal of generic evaluations in LTR-2 [NEDC-
31984P, ``Generic Evaluations of General Electric Boiling Water
Reactor Power Uprate'' and Supplement 1] to determine: which NRC and
industry generic communications were applicable to power uprate and
how they should be treated; analytical evaluations that could be
generically approved; bounding evaluations of components and
equipment, and; the effect of power uprate on safety margin. These
generic evaluations are supplemented by plant specific evaluations.
The Power Uprate Safety Analysis Report (PUSAR) describes the
dependence placed on References 1 and 2, the additional analyses
that were performed, the results of these additional analyses and
overall conclusions on the safety impacts of power uprate.
The plant systems and components will be within design limits at
power uprate conditions with minor modifications. At uprated power,
the power plant will not be operated in a manner that is different
from current operations except for limited changes to operating
parameters such as primary system pressure, steam flow and feedwater
temperature. Setpoints are revised as necessary to reflect new
operational conditions and analyses. The ECCS-LOCA [emergency core
cooling system-loss-of-coolant accident] analysis using current
practices demonstrates compliance with design and regulatory
acceptance criteria at uprated power.
The radiological consequences of accidents have been evaluated
using more current methodologies with consistent assumptions and
continue to meet acceptance criteria. Compliance with NRC dose
criteria using current methodologies is discussed in Section 9.2 of
the PUSAR. The effect of power uprate on dose analyses now discussed
in the FSAR [Final Safety Analysis Report] were qualitatively
assessed recognizing that power uprate increases doses in direct
proportion to the 4.1% increase in thermal power. An increase of
4.1% to the calculated doses currently identified in FSAR Chapter 14
indicates that a reevaluation using the original methodology would
have demonstrated compliance with current NRC dose criteria. A
review of Table 14.4-2 indicates that, with the 4.1% increase,
offsite doses would be substantially less than NRC allowable values.
A review of Table 14.8-1 indicates that, with the 4.1% increase,
control room doses would be substantially less than NRC allowables
except for the main steam line break (MSLB). However, the MSLB dose
would drop well below allowables once the proposed change on
allowable coolant activity (reduces the limit by more than a factor
of ten) is accounted for.
[Thus, based on the above analysis and supplemental analyses
performed by the NRC staff, the increase in power level discussed
herein and associated Technical Specification changes do not
significantly increase the probability or consequences of an
accident previously evaluated.]
2. create the possibility of a new or different kind of accident
from any accident previously evaluated.
Operation at uprated power involves no changes to the manner in
which the plant is operated. There are changes to operational
parameters and setpoints but analyses of these identified no new
failure modes or accident scenarios. The effects of transients and
accidents fall within design capabilities. Systems and components
are capable of operating and performing their safety functions at
uprated power. No mechanisms for creating a new or different
accident were identified.
3. involve a significant reduction in a margin of safety.
The power uprate will not result in significant increases to
primary system temperature and pressure due to postulated operating
transients or accidents. These and other margins of safety have been
discussed in the PUSAR, where it is demonstrated that there will be
no reductions in the margin of safety because the plant will still
meet its design and regulatory acceptance criteria. For example, the
core will continue to be operated with the same margin to the safety
limit minimum critical power ratio. Fuel thermal limits will
continue to meet NRC acceptance criteria. Plant systems and
equipment are designed for uprated power conditions and have been
evaluated for their capability to perform at uprated conditions.
They will continue to perform within design limits.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: December 22, 1993
Description of amendment request: The proposed amendment to the
James A. FitzPatrick Technical Specifications proposes to remove the
reference to American Society for Testing and Materials (ASTM) Standard
D 270-65 from Surveillance Requirement 4.12.A.1.i. ASTM D 270-65, which
specifies procedures to draw a representative fuel oil sample, has been
superseded and is no longer in effect. The FitzPatrick Surveillance
Procedure currently requires sampling in accordance with ASTM D 270-65
but notes that it has been superseded by ASTM D 4057-88. The proposed
change makes Surveillance Requirement 4.12.A.1.i consistent with fuel
oil Surveillance Requirement 4.9.C.1, for the emergency diesel
generators (EDGs), by adopting the current industry fuel oil sampling
standard. There are no changes to the acceptance criteria for fuel oil
quality which are based on ASTM D 975-81.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
Plant accident analyses are not affected by the Technical
Specification change. The change removes reference to ASTM D 270-65
as the method for obtaining samples of diesel fuel oil from
Surveillance Requirement 4.12.A.1.i. The American Society for
Testing and Materials has superseded ASTM D 270-65 with ASTM D 4057.
The proposed change provides improved flexibility to adopt standards
as they are issued without requiring a license amendment and makes
Surveillance Requirement 4.12.A.1.i consistent with both
Surveillance Requirement 4.9.C.1 for the EDG and the STS [Standard
Technical Specification]. The nature of this change will not cause
any increase in the probability or consequences of previously
evaluated accidents.
2. create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed change involves no hardware modifications to any
plant structures, systems or components. The change removes
reference to ASTM D 270-65 as the method for obtaining samples of
diesel fuel oil from Surveillance Requirement 4.12.A.1.i. The
American Society for Testing and Materials has superseded ASTM D
270-65 with ASTM D 4057. The nature of this change is such that no
new or different kind of accident can be created.
3. involve a significant reduction in the margin of safety.
The proposed change will not cause a reduction in the margin of
safety. The results of the plant accident analyses continue to bound
operation under the proposed changes so there is no reduction in the
margin of safety. The change removes reference to ASTM D 270-65 to
allow the use of ASTM D 4057-88 as the method for obtaining samples
of diesel fire pump fuel oil. The change will make Surveillance
Requirement 4.12.A.1.i consistent with existing Surveillance
Requirement 4.9.C.1 for the EDG and allow current standards to be
used for fuel oil sampling. Revisions of this nature will not cause
a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: December 28, 1993
Description of amendment request: The proposed amendment to the
James A. FitzPatrick Technical Specifications (TSs) revises the scope
of startup and power escalation reports to reflect the guidance
provided by the Standard Technical Specifications (STS). Current TS
requirements state, ``The report shall address each of the tests
identified in the FSAR [Final Safety Analysis Report] ....''
The list of tests provided in the FSAR includes many tests which
were required for the initial plant startup but are not performed for
subsequent startups. Currently, the licensee satisfies TS requirements
in the startup reports by identifying tests which are not performed,
their purpose, and the reason for not performing them.
The inclusion of such a section in the startup reports is
unnecessary. This amendment would revise the TSs to permit subsequent
startup reports to address only those tests that are actually
performed.
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes involve no hardware modifications, no
changes to the operation of any system or component, no changes to
structures, and alters future startup reports by addressing only
tests that are performed. A change in reporting format will not
eliminate the performance of startup tests that are necessary or
required. These changes can not alter the probability or
consequences of existing accident analyses as documented in the FSAR
or the NRC staff SER [Safety Evaluation Report].
2. create the possibility of a new or different kind of accident
from those previously evaluated.
The changes do not alter the testing procedures or
methodologies. The changes provide a basis, in accordance with NRC
guidance, for allowing future startup test reports to mention and
discuss only those tests that are actually performed. These changes
can not result in a new or different type of accident than those
previously evaluated.
3. involve a significant reduction in the margin of safety.
There are no changes to tests that are performed in support of
plant startup [***]. Removing a requirement for reporting on tests
that are not performed will not cause a reduction in any margin of
safety.
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 NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: December 29, 1993
Description of amendment request: The proposed amendment to the
James A. FitzPatrick Technical Specifications would eliminate an
inconsistency between the Reactor Coolant System (RSC) leakage
detection and operability requirements in Limiting Conditions for
Operation (LCO) 3.6.D.1 and 3.6.D.4.
LCO 3.6.D.1 currently requires that RCS leakage be below specified
limits when there is irradiated fuel in the reactor vessel and reactor
coolant temperature is greater than 212 deg.F. LCO 3.6.D.4 requires the
leakage monitoring systems to be operable during power operation (i.e.,
when the mode switch is in the Startup/Hot Standby position or the Run
position with the reactor critical above 1 percent rated power, as
defined per Specification 1.0.O). These two LCOs are not consistent.
The proposed revision of LCO 3.6.D.4 will take the more conservative
approach of requiring the leakage monitoring systems to be operable
when the leakage limits of LCO 3.6.D.1 are in effect.
The proposed changes also make editorial corrections which are
considered administrative in nature.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment involves no hardware changes, no changes
to the operation of any systems or components and no changes to
structures. It alters an LCO to require plant leakage detection
systems to be operable during the same plant conditions that RCS
leakage limits apply. The revised LCO requires leakage detection
systems in operational modes consistent with other portions of the
Technical Specifications. Additional changes include editorial
corrections such as correct specification numbering, proper system
identification, and clarification of a surveillance requirement
cross reference. Since the change to the LCO for leakage detection
will require operability under a greater range of plant conditions
to be consistent with detection requirements, there is no change to
previously determined accident probabilities or consequences. The
editorial changes have no adverse safety implications.
2. create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed amendment involves no hardware changes, no changes
to the operation of any systems or components and no changes to
structures. It alters the Technical Specifications only to the
extent of making two LCOs consistent by requiring the leakage
detection system to be operable when leakage limits apply and making
editorial changes. Editorial changes and increasing the plant
conditions for leakage system operability can not create the
possibility of a new or different kind of accident from those
previously evaluated since the editorial changes have no safety
significance and the operability changes are being made for
consistency with the modes when leakage detection is required to
function.
3. involve a significant reduction in the margin of safety.
The proposed amendment revisions involve no hardware changes, no
changes to the operation of any systems and no changes to
structures. The revised LCO criteria for RCS leakage detection
system operability has increased the plant conditions when
operability is required to match the plant conditions when leakage
limits apply. Editorial changes and expanded operability
requirements in the LCOs will not result in any change to existing
safety margins.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: December 29, 1993
Description of amendment request: This proposed amendment to the
James A. FitzPatrick Technical Specifications, Appendix B, revises
Surveillance Requirement 3.1.a and Table 3.10-2. The existing
Surveillance Requirement in Specification 3.1.a, which references Table
3.10-2, requires the performance of instrument checks, tests, and
calibrations to assure the operability of specific gaseous effluent
radiation monitors. It does not require any surveillances for the
associated data recorders. However, Limiting Condition for Operation
(LCO) 3.1.a specifically states that ''...pathways shall be monitored
and recorded ....'' which requires the recorders to be operable for the
gaseous effluent monitoring system to be considered operable. The
proposed changes to the Radiological Effluent Technical Specifications
(RETS) add the radiation monitor recorders to the Surveillance
Requirement and Table 3.10-2 for the gaseous effluent monitoring
system. This will provide the surveillance requirements for the data
recorders.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment involves no hardware changes, no changes
to the operation of any systems or components and no changes to
structures. It alters procedures by including the recorders in the
Surveillance Requirements for determining the operability of the
radiological monitoring/recording instrumentation. The inclusion of
the data recorders in the Surveillance Requirements for checking,
testing, and calibration does not reduce gaseous effluent monitoring
capability while providing greater assurance that data is recorded.
Adding the existing system surveillance to the Technical
Specifications will not have any affect on previously evaluated
accident probabilities or consequences.
2. create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed amendment involves no hardware changes, no changes
to the operation of any systems or components and no changes to
structures. It alters procedures by including the data recorders in
the Surveillance Requirements for the radiation monitors. These
changes do not affect the manner in which the gaseous effluent
radiation monitoring system is operated or tested. The resulting
changes do not pose a safety issue concern different from those
analyzed previously in the FSAR [Final Safety Analysis Report] or
the NRC staff SER [Safety Evaluation Report].
3. involve a significant reduction in the margin of safety.
The proposed amendment involves no hardware changes, no changes
to the operation of any systems and no changes to structures. It
alters procedures by including the data recorders in the
Surveillance Requirements for the radiation monitors. The revised
surveillance requirement increases the scope of surveillance for the
gaseous effluent radiation monitoring system, for consistency with
the LCO. It does not cause any reduction in any safety margins.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: January 11, 1994
Description of amendment request: This proposed amendment to the
James A. FitzPatrick Technical Specifications (TS) would temporarily
waive the 2-year maximum surveillance interval for the Type C test
(local leak rate test) of the shutdown cooling isolation valves (10MOV-
17 and 10MOV-18). The waiver would permit deferring this test until the
next refueling outage, currently scheduled for the end of November
1994.
TS 4.7.A.2.e(5) and 10 CFR Part 50 Appendix J require the
containment isolation valves to be leak-rate tested during each
refueling outage, but at intervals of not greater than 2 years. As a
result of the extended 1991-1993 refueling outage, and the length of
the current operating cycle, tests of all containment isolation valves
are due prior to the start of the next refueling outage. Two mid-cycle
outages were scheduled, in part, to accommodate the testing of these
valves, except for the shutdown cooling isolation valves. The shutdown
cooling isolation valves cannot be removed from service during a non-
refueling outage to accommodate a leak-rate test since its associated
system is needed to remove reactor decay heat. The reliability of the
isolation valve design, and the very low probability that the shutdown
cooling system penetration would result in a pathway for leakage to the
reactor building, as discussed in the application, is provided as
justification for this one-time schedular extension.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change is limited to a one-time schedular extension
in the shutdown cooling isolation valve Type C test. The change does
not introduce any new modes of plant operation, make any physical
changes, or alter any operational setpoints. Therefore, the change
does not degrade the performance of any safety system assumed to
function in the accident analysis. The extension will not result in
a significant increase in valve leakage considering that: (1) the
valves are infrequently stroked, and then only when the reactor
vessel is at low pressure, (2) monitoring the normal operating
status of the RHR [residual heat removal] system assures the absence
of gross valve leakage, and (3) the valves were replaced with valves
of an improved design which has been confirmed by past Type C tests
to exhibit satisfactory leak rate performance. For these reasons,
the change does not involve a significant increase in the
probability of an accident.
The change does not involve a significant increase in the
consequences of an accident evaluated since any leakage through the
shutdown cooling penetration will not significantly increase for
reasons discussed in the previous paragraph, and such leakage is
negligible compared to the main steam line break accident analyzed
in the FSAR [final safety analysis report].
2. create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed change does not introduce new accident initiators
or failure mechanisms since the change does not alter the physical
characteristics of any plant system or component. The change is
limited to a one-time schedule extension for the shutdown cooling
isolation valve Type C tests.
3. involve a significant reduction in the margin of safety.
There is a very low probability of a significant increase in
valve leakage considering the demonstrated reliability of the
current valve design, the infrequent use of the valves, and the
monitoring of the normal operating status of the RHR system.
Moreover, any potential incremental benefit of performing the tests
within the two year requirement would not be sufficient to offset
the increased occupational radiation exposure associated with
testing, and the risk to plant safety associated with the removal
from service of the primary method of decay heat removal.
Consequently, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 1, 1993 TS 93-09
Description of amendment request: The proposed changes would revise
the setpoints and time delays for the auxiliary feedwater (AFW) and 6.9
kv shutdown board loss of voltage and degraded voltage instrumentation.
The proposed changes would affect Technical Specification (TS) Tables
3.3-3, 3.3-4, 3.3-5, 4.3-2, and the Bases for Specification 3/4.7.1.2.
Table 3.3-3 would be revised: (1) to reflect the use of a two-out-of-
three voltage sensor logic for loss of power detection (and one-out-of-
two logic scheme for the timing relays) by adding requirements for the
respective AFW, 6.9 kv shutdown board, and emergency diesel generator
(EDG) voltage sensors and load shed timers; (2) by changing the
description of the functional unit from ``Station Blackout'' to ``Loss
of Power Start;'' (3) by adding a footnote to indicate that the new
requirements apply only to the shutdown boards on the same unit; (4) by
increasing the modes for which the EDG sensors and timers must be
operable to include Modes 5 and 6 when the associated EDG must be
operable; (5) by changing the associated Action statements to reflect
the design changes and consistency; (6) to reflect consistent
terminology in the Table and action statements; (7) by changing the
minimum number of channels that would be required to be operable; (8)
by changing the footnote to reflect the conditions when the loss-of-
power instrumentation is required to be operable in Modes 5 and 6; and
(9) by adding an exclusion to Specification 3.0.4. Similar changes were
proposed for Tables 3.3-4, 3.3-5 and 4.3-2. The proposed change to
Bases 3/4.7.1.2 would clarify AFW operability on loss of power.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revision supports the implementation of design
logic and setpoint changes to the loss-of-power relaying. This
relaying is designed to ensure adequate voltage is available to
safety-related loads in order to enhance their operability and
support accident mitigation functions and to provide for auxiliary
feedwater (AFW) pump starts. The design changes alter relay logic
and delete unnecessary relaying, but do not change the diesel
generator (D/G) start and load-shedding actuations that result from
loss-of-power conditions. Therefore, no new actuations or functions
have been created; and because the existing and proposed functions
provide for accident mitigation considerations that are not the
source of an accident, the probability of an accident is not
increased. The deletion of the 6.9-kilovolt shutdown board normal-
feeder undervoltage relays actually reduces the potential for
inadvertent shutdown board blackouts as a result of short-duration
voltage transients or instrument failures.
The setpoints and time delays for loss-of-power functions have
been modified based on the guidelines developed by the Electrical
Distribution System Clearinghouse as evaluated and determined
through detailed analysis by TVA. This design is documented in TVA
Calculations SQN-EEB-MS-TI06-0008, 27DAT, and DS-1-2 and is
available for NRC review at the SQN site. The assigned values are
conservative settings that will ensure adequate voltage is supplied
to safety-related loads for accident mitigation and safety functions
under normal, degraded, and loss-of-offsite-power voltage conditions
with appropriate time delays to prevent damage to electrical loads
and minimize premature or unnecessary actuations. The identification
of loss-of-voltage conditions is enhanced by the design changes to
ensure the timely sequencing of loads onto the D/G and the
initiation of AFW pump starts for accident mitigation. Because there
are no reductions in safety functions resulting from the design
logic, setpoint, and time-delay changes to the loss-of-power
instrumentation and offsite dose levels for postulated accidents
will not be increased, the consequences of an accident are not
increased.
The applicable mode addition, TS 3.0.4 exclusion deletion, and
response time measurement clarification incorporated in the proposed
change do not affect plant functions. These changes reflect the
requirements that SQN has been maintaining and serve to clarify the
requirements to provide consistency of application and easier
understanding. The AFW footnote addition and bases revision only
clarify operability conditions that are consistent with the plant
design for the AFW pump and loss-of-power instrumentation. Because
there are no changes to plant functions or operations, these
revisions have no impact on accident probabilities or consequences.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
As described above, the loss-of-power instrumentation ensures
adequate voltage to safety-related loads by initiating D/G starts
and load shedding and provides for AFW pump starting, but is not
considered to be the source of an accident. Although the design
logic, setpoint, and time-delay actuation criteria have changed, the
output functions to various plant systems that actuate for load
shedding and D/G starts remain the same. Therefore, actuation
criteria have been affected, but not safety functions, and the TVA
evaluation has confirmed that the new design enhances the ability to
maintain adequate voltage to support safety functions. Since safety
functions have not changed and the new loss-of-power instrumentation
design continues to support operability of safety-related equipment,
no new or different accident is created.
The applicable mode addition, TS 3.0.4 exclusion deletion, and
response time measurement clarification, as well as the AFW
operability clarifications, do not affect plant functions and will
not create a new accident.
3. Involve a significant reduction in a margin of safety.
The proposed loss-of-power TS changes support design logic,
setpoint, and time-delay requirements that have been verified by TVA
analysis to provide acceptable voltage levels for safety-related
components. In determining the acceptability of these voltage
levels, the minimum voltage for operation as well as detrimental
component heating resulting from sustained degraded-voltage
conditions were considered. This design ensures that safety-related
loads will be available and operable for normal and accident plant
conditions. The applicable mode addition, TS 3.0.4 exclusion
deletion, response time measurement clarification, and AFW
operability clarifications provide enhancements to TS requirements
and do not affect plant functions. Therefore, no safety functions
are reduced by these changes and there is no reduction in the margin
of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: November 18, 1993 (TS 93-17)
Description of amendment request: The proposed change would delete
the requirement of License Condition 2.H of the Operating License. This
license condition requires reporting the violation of certain license
conditions to the NRC Regional Administrator within 24 hours by phone
and facsimile, and a followup report within 14 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change is a deletion of an administrative reporting
requirement that does not in any way affect a previously analyzed
accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Since there are no changes in the way the plant is operated, the
potential for an unanalyzed accident is not created. The proposed
change is administrative in nature and does not affect any accident
initiators for SQN. No new failure modes are introduced.
3.Involve a significant reduction in a margin of safety.
Plant safety margins are established through limiting conditions
of operation, limiting safety system settings, and safety limits
specified in the TSs. As a result of the proposed amendment, there
will be no changes to either the physical design of the plant or to
any of these settings and limits. The proposed changes are
administrative and do not affect the safe operation of SQN.
Therefore, there will be no changes to any of the margins of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: December 7, 1993 (TS 93-16)
Description of amendment request: The proposed change would
incorporate the following changes to Section 6.0, ``Administrative
Controls,'' of the Technical Specifications (TS): (1) delete Section
6.1.2 that indicates that the Corporate Manager of Radiological Control
has the responsibility for the radiological environmental program, dose
calculations, and projections; (2) added a requirement in Section
6.5.1.6 for review of the Offsite Dose Calculations Manual (ODCM) by
the Plant Operations Review Committee (PORC); (3) added implementation
of the ODCM to Section 6.8.1 as an activity requiring written
procedures and deleted the requirement that they be maintained by the
Radiological Control Group; (4) delete the requirement in Section
6.8.4.a for Radiological Control to implement and control the ODCM; (5)
changed the review authority for changes to the ODCM from the
Radiological Assessment Review Committee (RARC) to the PORC and added a
reference to Specification 6.5.1A in Section 6.14.1.2; (6) approval
authority for deviations from the overtime guidelines in Section
6.2.2.g would be changed to show that Plant Manager designee also has
the authority; (7) PORC member titles would be changed in accordance
with the current organizational structure; (8) the requirements for the
RARC would be deleted from Sections 6.5.2.7.i, 6.5.3, 6.10.2.k and the
index; (9) move the condenser inleakage monitoring requirement from
Section 6.8.5.c.(vii) to 6.8.5.c.(iii) and delete the requirement to
repair, plug, or isolate leaks; (10) change the title ``Shift
Supervisor'' (SS) to ``Shift Operations Supervisor'' (SOS) in various
locations in Section 6 and in Operating License Items 2.C.(23).A and
2.C.(16).a for Units 1 and 2 respectively; (11) change the title of the
Senior Vice President, Nuclear Group, in Section 6.2.1 to Senior Vice
President, Nuclear Power; (12) change the title of the Operational
Quality Assurance Program in Section 6.5.2.8.d to Nuclear Quality
Assurance Program; and (13) move the requirement for implementation of
the Quality Assurance Program for environmental monitoring from Section
6.8.4.b to Section 6.8.1.h. Other administrative changes related to
these changes were also submitted that affect Section 6.0, the
operating license, index, and definitions.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). The operation of Sequoyah Nuclear Plant (SQN) in
accordance with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes only affect the administrative controls
found in Section 6.0 of the SQN TSs and the operating license. No
plant equipment or operating practices are affected by these
changes. The revised administrative controls will continue to
adequately implement administrative activities to support plant
nuclear safety. Since there are no physical changes to the plant,
there is no increase in the probability of an accident because these
administrative controls are not the source of previously evaluated
accidents. Similarly, with no change to plant equipment or operating
requirements, the plant response to accident conditions and
therefore the consequences of an accident remain unchanged. These
proposed changes will not increase the consequences of an accident
and offsite dose rates will not be impacted.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The administrative controls affected by the proposed changes are
not considered to be the source of any accident and these changes
will not alter any plant features or processes. Therefore, the
proposed changes will not create the possibility of a new or
different kind of accident and the administrative controls will
continue to implement the actions necessary to support plant
activities and nuclear safety.
3. Involve a significant reduction in a margin of safety.
Plant features and setpoints remain unchanged by the proposed
changes to the administrative controls. The margins of safety
established by the SQN design are not affected by these changes. The
proposed administrative controls will continue to maintain the
actions and programs that ensure appropriate plant design,
operation, and procedures to support the required margin of safety.
Therefore, the proposed changes will not reduce the margin of
safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: September 29, 1993
Description of amendment request: The proposed amendments would
change the test intervals from monthly to quarterly, consistent with
the Inservice Test Program, for several pumps and related systems,
including safety injection, residual heat removal and containment spray
pumps.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendments will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The addition of specific general considerations related to
equipment surveillance requirements and their relationship to
equipment operability are administrative in nature. They do not
change the interpretation or intent of the Technical Specifications.
These conditions are consistent with the Westinghouse Standard
Technical Specifications (STS).
The addition of the specific requirement for the Inservice Test
Program to Section 15.4.2 of the Technical Specifications only
reiterates the requirements of 10 CFR 50.55a(g). This requirement
does not implement any new requirements on the operation or testing
of equipment.
The decrease in the number of equipment operational transients
due to the increase in the surveillance interval for the Safety
Injection System (SI), Residual Heat Removal (RHR) System, and
Containment Spray (CS) System pumps and valves will result in an
increase in system availability. Reduced testing is also expected to
have a positive affect on overall equipment reliability since
frequent testing results in increased wear and potential for
equipment failure. Other actions including a monthly verification of
system lineups for the SI, CS and RHR systems provides increased
assurance of system operability between surveillance tests. The
potential for equipment problems to go undetected for a longer
period of time is small as indicated by equipment surveillance
history.
Therefore, these changes will not effect the probability or
consequences of previously analyzed accidents.
2. The proposed amendments will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
These changes only affect the equipment testing frequency.
Equipment design, operation and the methods of testing will not be
changed. Therefore, the proposed changes cannot create the
possibility of a new or different kind of accident than any accident
previously evaluated.
3. The proposed amendments will not involve a significant
reduction in the margin of safety.
The proposed changes which implement the requirements of 10 CFR
50.55a(g) and clarify the general considerations related to
equipment surveillances are administrative in nature. They do not
change the intent of any existing license or other requirement.
The increase in equipment surveillance intervals will result in
an improvement in equipment and system availability and reliability.
Surveillance of equipment will be performed as required by the
regulations and Section XI of the ASME Boiler and Pressure Vessel
Code. These proposed changes will reduce the potential for equipment
failures due to unnecessary testing.
Adequate assurance is provided by testing in accordance with the
ASME Code requirements and periodic verification of system lineups
to ensure that the affected systems remain operable and capable of
performing their design function. Therefore, a reduction in a margin
of safety will not occur.
The NRC staff has reviewed the licensee's analysis and based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: December 1, 1993
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TS) by incorporating technical and administrative changes to TS 3.10,
Control Rod and Power Distribution Limits. The proposed changes
eliminate specifications for fuel designs no longer used at Kewaunee,
specify required actions to be taken upon exceeding control bank
insertion limits, and revise the limits for Departure from Nucleate
Boiling (DNB) related parameters to assure operation within the
assumptions of the Updated Safety Analysis Report (USAR) analyses.
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) TS 3.10.b.1, 3.10.b.4 and Table TS 3.10-2
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes eliminate the specifications for fuel that
is no longer used at the Kewaunee Nuclear Power Plant. Eliminating
these specifications does not affect the probability of any accident
previously evaluated.
The specifications for the current fuel vendor are being
retained and ensure the consequences of previously evaluated
Departure from Nucleate Boiling (DNB) related accidents are
enveloped by the Updated Safety Analysis Report (USAR) analyses.
Therefore, these changes will not increase the consequences of an
accident previously evaluated in the USAR.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not alter the plant configuration,
operating setpoints, or overall plant performance. Therefore, it
does not create the possibility of a new or different kind of
accident.
3) Involve a significant reduction in the margin of safety.
The proposed change deletes specifications for fuel that is no
longer used at the Kewaunee Nuclear Power Plant. The limits for the
current fuel vendor are retained and are not affected by this
proposed change. This does not alter the input or assumptions of the
safety analysis, and therefore it will not involve a reduction in
the margin of safety.
(b) TS 3.10.d
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change specifies the required actions to be taken
when the control bank insertion limits are exceeded. The current TS
requires compliance with the control bank insertion limits, but
gives no corrective action for when these limits are exceeded. The
new specification requires operators to initiate boration to restore
shutdown margin within one hour of exceeding the control bank
insertion limits and to restore the control banks to within the
limits within 2 hours. If either of these requirements cannot be
achieved, within 1 hour the operators must initiate actions to
achieve Hot Standby within the next 6 hours and Hot Shutdown within
the following 6 hours. Adding these requirements clarifies and
enhances the Technical Specifications and will have no impact on the
probability of an accident previously evaluated.
The proposed addition is conservative and ensures that proper
and adequate measures are taken when the control bank exceeds the
control bank insertion limits. Therefore, this addition will not
increase the consequences of an accident previously evaluated in the
USAR.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not alter the plant configuration,
operating setpoints, or overall plant performance. Therefore, it
does not create the possibility of a new or different kind of
accident.
3) Involve a significant reduction in the margin of safety.
The proposed change is conservative and clarifies the necessary
actions to be taken when control bank insertion limits are exceeded.
This proposed change is an enhancement to the specification and does
not reduce the margin of safety.
(c) TS 3.10.k
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change decreases the maximum RCS inlet temperature
limit for steady state 100% operation from 536.5 degrees Fahrenheit
to 535.5 degrees Fahrenheit. The value of 539.5 degrees Fahrenheit
is the assumed RCS inlet temperature for the DNB related accidents
analyzed in the USAR. These accidents are the Uncontrolled Rod
Cluster Control Assembly Withdrawal at Power Accident, the
Uncontrolled Rod Cluster Control Assembly Withdrawal from a
Subcritical Condition Accident, the Rod Cluster Control Assembly
Misalignment, the Start-Up of an Inactive Reactor Coolant Loop
Accident, the Excessive Heat Removal due to Feedwater System
Malfunction Accident, the Excessive Load Increase Incident, the Loss
of Reactor Coolant Flow Accident, the Loss of External Electrical
Load, the Steam Line Break, and the Rod Cluster Control Assembly
Ejection. A four degree assumed instrument error reduces the maximum
allowed RCS inlet temperature to 535.5 degrees Fahrenheit.
Decreasing this value to ensure consistency with the USAR analysis
assumptions will have no impact on the probability of an accident
previously evaluated.
The proposed change is conservative to ensure that the
consequences of a previously evaluated DNB-related accident are
enveloped by the USAR analysis. Therefore, this change will not
increase the consequences of an accident previously evaluated in the
USAR.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not alter the plant configuration,
operating setpoints, or overall plant performance. Therefore, it
does not create the possibility of a new or different kind of
accident.
3) Involve a significant reduction in the margin of safety.
The proposed change is conservative and is consistent with the
assumptions in the USAR. This proposed change is an enhancement to
the specification and does not reduce the margin of safety.
(d) TS 3.10.l
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change increased the minimum RCS pressure limit for
steady-state 100% power operation from the currently specified 2200
psig to 2205 psig. The value of 2205 psig (30 psig below the nominal
design value of 2235 psig) was the assumed initial condition for the
DNB related accidents analyzed in the USAR. These accidents are the
Uncontrolled Rod Cluster Control Assembly Withdrawal at Power
Accident, the Uncontrolled Rod Cluster Control Assembly Withdrawal
from a Subcritical Condition Accident, the Rod Cluster Control
Assembly Misalignment, the Start-Up of an Inactive Reactor Coolant
Loop Accident, the Excessive Heat Removal due to Feedwater System
Malfunction Accident, the Excessive Load Increase Incident, the Loss
of Reactor Coolant Flow Accident, the Loss of External Electrical
Load, the Steam Line Break, and the Rod Cluster Control Assembly
Ejection. Increasing this value to ensure consistency with the USAR
analysis assumptions will have no impact on the probability of an
accident previously evaluated.
The proposed change is conservative to ensure that the
consequences of a previously evaluated DNB-related accident is
enveloped by the USAR analysis. Therefore, this change will not
increase the consequences of an accident previously evaluated in the
USAR.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not alter the plant configuration,
operating setpoints, or overall plant performance. Therefore, it
does not create the possibility of a new or different kind of
accident.
3) Involve a significant reduction in the margin of safety.
The proposed change is conservative and is consistent with the
assumptions in the USAR. This proposed change is an enhancement to
the specification and does not reduce the margin of safety.
(e) TS 3.10.m
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1)Involve a significant increase in the probability or
consequences of an accident previously evaluated.
TS 3.10.m.1 provides the limits and required actions to be taken
when the RCS flow rate per loop is less than the USAR analysis
assumptions. Decreasing the flow limit to be consistent with the
USAR assumptions will have no impact on the probability of an
accident previously evaluated. Compliance with the flow limit
assumed in the USAR analyses ensures the consequences of previously
evaluated DNB-related accidents are enveloped by the USAR analyses.
Therefore, this change will not increase the consequences of an
accident previously evaluated in the USAR.
The proposed revision places an additional restriction on RCS
flow when less than the design flow rate of 89,000 gallons per
minute per loop. The intent of specifying action in accordance with
3.10.n is to ensure that reactor power is reduced to a point at
which the DNB ratio margin is restored. Compliance with this
specification will not increase the probability of an accident
previously evaluated, nor increase the consequences of an accident.
The intent of TS 3.10.m.2 is to clarify the conditions under
which the reactor coolant flow rate is verified. The conditions,
methodology, and uncertainties associated with this verification are
not changed by this specification. Clarifying the TS by inclusion of
the conditions for verifying the flow rate will not increase the
probability or consequences of any accident previously evaluated.
2) Create the possibility of a new or different kind of accident
from any accidents previously evaluated.
A new or different kind of accident from those previously
evaluated will not be created by this TS change. The proposed
amendment does not alter the plant configuration, operating
setpoints or overall plant performance.
3) Involve a significant reduction in the margin of safety.
A specification on steady-state RCS flow rate is necessary to
ensure DNB ratio criteria will be met during the DNBR limiting
events analyzed in the USAR. Reducing the reactor coolant flow limit
to the value assumed in the USAR analyses does not result in a
reduction in the margin of safety.
The additional restrictions being imposed if the specified limit
is not met provide additional assurance the DNBR margin will be
restored. These additional restrictions do not exist in the current
TS. The imposition of these restrictions results in an enhancement
to the margin of safety.
Clarifying the TS by inclusion of conditions under which the
flow verification is to be performed will not reduce the margin of
safety. Existing approved constraints, methodology, and
uncertainties are not being changed by this clarification.
(f) TS 3.10.n
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The intent of this new TS is to outline the actions required
when the limits of TS 3.10.k (RCS temperature), TS 3.10.l (RCS
pressure) and TS 3.10.m.1 (RCS flow) are exceeded. Collectively,
these three specifications place limits on the DNB-related
parameters to assure each is maintained within the normal steady-
state envelope assumed in the USAR safety analysis. This
specification is an enhancement to our existing specification to add
clear guidance which does not presently exist. Providing this
information for the plant staff and operators will not increase the
probability of an accident previously evaluated.
The addition of this action statement will ensure that the
consequences of an analyzed accident are not increased. The proposed
specification allows 2 hours to evaluate and restore parameters to
within limits. If this time frame is not satisfied, then within the
next 6 hours, power is reduced in order to restore a margin of
safety. Following analysis, thermal power may be raised not to
exceed a level analyzed to maintain a minimum DNBR of 1.30.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed amendment does not alter the plant configuration,
operating setpoints or overall plant performance.
3) Involve a significant reduction in the margin of safety.
Addition of the specification is an enhancement to the current
specification which does not alter input to the safety analysis.
Therefore, it will not involve a reduction in the margin of safety.
(g) Administrative changes to Section TS 3.10 including Figure
TS 3.10-2.
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated, or
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3) Involve a significant reduction in the margin of safety.
The proposed changes are administrative in nature and do not
alter the intent or interpretation of the TS. Therefore, no
significant hazards exist.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: John N. Hannon
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: December 7, 1993
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
5.3.a.1 to provide flexibility in the repair of fuel assemblies
containing damaged and leaking fuel rods by reconstituting the
assemblies, provided that an NRC-approved methodology is used. This
proposed change is consistent with guidance provided in Supplement 1 to
Generic Letter (GL) 90-02, ``Alternative Requirements for Fuel
Assemblies in the Design Features Section of Technical
Specifications,'' dated July 21, 1992. In addition, administrative
changes to KNPP TS Section 5 have been proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
The proposed changes were revised in accordance with the
provision of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1) involve a significant increase in the probability or
consequences of an accident previously evaluated.
This proposed change to the requirements for ``Fuel Assemblies''
in the ``Design Features'' section of the KNPP TS will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. This proposed change will not result
in modifications to fuel assemblies that would have a significant
effect on safety because of the requirement to implement these
changes using an NRC-approved methodology. This requirement will
confirm conformance to existing design limits and confirm that
safety analyses criteria are met before operation during the next
fuel cycle. This license amendment request is consistent with
guidance provided by the NRC and will result in flexibility for
improved fuel performance.
2) create the possibility of a new or different kind of accident
from any accident previously evaluated.
The creation of new or different kind of accident from any
previously evaluated accident is not considered a possibility
because the changes are administrative in nature and do not
represent an actual modification to the plant or change its safety
analyses.
3) involve a significant reduction in the margin of safety.
The margin of safety is maintained by adherence to other fuel
related TS limits and the USAR design bases. The changes do not
directly affect any safety system or the safety limits, and thus
does not affect the plant margin of safety.
Accordingly, these proposed changes do not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: John N. Hannon
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of application for amendments: October 19, 1993
Brief description of amendments: The amendments change the
Technical Specifications (TS) add a footnote to TS 4.6.1.2.b that
allows a one-time exemption from the accelerated containment integrated
leak rate test (CILRT) requirements to return the CILRT frequency for
both units to a normal Type A test interval.
Date of issuance: January 11, 1994
Effective date: January 11, 1994
Amendment Nos.: 167 and 198
Facility Operating License Nos. DPR-71 and DPR-62. Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59745) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 11, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
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 application of amendments: May 3, 1993, as supplemented
August 11, 1993.
Brief description of amendments: The amendments revise the limiting
conditions for operation and surveillance requirements related to the
Low Pressure Service Water System.
Date of issuance: January 13, 1994
Effective date: To be implemented within 30 days from the date of
issuance.
Amendment Nos.: 203, 203, and 200
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 13, 1993 (58 FR
52983) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 13, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear
One,Unit No. 2, Pope County, Arkansas
Date of application for amendment: October 27, 1993
Brief description of amendment: The amendment relocated the
requirement of Technical Specification 4.5.2.g.1 to verify the correct
position of each electrical and/or mechanical position stop for the
Emergency Core Cooling System throttle valves within 4 hours of each
valve stroking operation or maintenance on the valve, to procedures
that control the maintenance and operation of these valves.
Date of issuance: January 14, 1994
Effective date: To be implemented within 30 days from the date of
issuance.
Amendment No.: 155
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64606) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 14, 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: August 11, 1993
Brief description of amendment: The amendment changed the Grand
Gulf Nuclear Station Technical Specifications to support compliance
with the new requirements of Title 10 Code of Federal Regulations Part
20 and Part 50.36a. The request to change the wording of TS 1.46 which
relates to the definition of an UNRESTRICTED AREA remains under
consideration and will be the subject of a future licensing action.
Date of issuance: January 10, 1994
Effective date: January 10, 1994
Amendment No: 111
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46233) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 10, 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.
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: August 27, 1993.
Description of amendment request: The amendment changes the
footnote on page 1 of License NPF-86 by deleting Vermont Electric
Generation and Transmission Cooperative, Inc., (Vermont), as one of the
entities for which North Atlantic Energy Service Corporation (North
Atlantic) is authorized to act. The change reflects the purchase of
Vermont's share of the Seabrook Station, Unit 1 by North Atlantic
Energy Corporation (NAEC) pursuant to a prior settlement of a claim by
Vermont against Public Service Company of New Hampshire (PSNH). NAEC
acquired PSNH's interest in the Seabrook Station, Unit 1 in accordance
with the Plan for Reorganization for PSNH.
Date of issuance: January 7, 1994
Effective date: To be implemented by May 30, 1994.
Amendment No.: 28
Facility Operating License No. NPF-86. Amendment revised the
License.
Date of initial notice in Federal Register: October 13, 1993 (58 FR
52990). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 7, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, New Hampshire 03833.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: October 7, 1992 as supplemented July 12,
1993
Brief description of amendment: The amendment changed the setpoint
limit for the degraded-voltage protection system referred to as the
offsite-power low signal.
Date of issuance: January 14, 1994
Effective date: January 14, 1994
Amendment No.: 159
Facility Operating License No. DPR-40. Amendment revised the
TechnicalSpecifications.
Date of initial notice in Federal Register: November 25, 1992 (57
FR 55584) The additional information contained in the supplemental
letter dated July 12, 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 January 14, 1994. No significant hazards consideration comments
received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: September 15, 1993
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) by implementing Generic Letters (GLs) 86-10 and
88-12. This removed the fire protection TSs and placed these parts in
the Updated Safety Analysis Report (USAR).
Date of issuance: January 14, 1994
Effective date: January 14, 1994
Amendment No.: 160
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59753) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 14, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Philadelphia Electric Company, Docket No. 50-352, Limerick
Generating Station, Unit 1, Montgomery County, Pennsylvania.
Date of application for amendment: August 3, 1993
Brief description of amendment: This amendment removes shutdown
system control valves and primary containment isolation valves from TS
Tables 3.3.7.4-1, ``Remote Shutdown Instrumentation and Controls,'' and
3.6.3-1, ``Primary Containment Isolation Valves,'' as a result of
eliminating the steam condensing mode of the Residual Heat Removal
system.
Date of issuance: January 12, 1994
Effective date: January 12, 1994
Amendment No. 65
Facility Operating License No. NPF-39. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50969) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 12, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: May 21, 1993, as supplemented
October 7, 1993, and December 3, 1993
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to incorporate the following changes:
(1) The safety injection system test frequency (specified in TS
Section 4.5.A.1.a) was changed to accommodate operation on a 24-month
cycle.
(2) The loss of normal AC in conjunction with a safety injection
signal test frequency (specified in TS Section 4.6.A.3) was changed to
accommodate operation on a 24-month cycle. This TS section was also
reformatted to improve clarity.
(3) The auxiliary feedwater system undervoltage automatic start
test frequency (specified in TS Table 4.1-1) was changed to accommodate
operation on a 24-month cycle.
(4) The auxiliary feedwater system main feedwater pump trip
automatic start test frequency (specified in TS Table 4.1-1) was
changed to accommodate operation on a 24-month cycle.
In addition, quarterly testing and 24-month calibration
requirements were added to TS Table 4.1-1 for the main steam line flow
instrumentation. These surveillances were added to ensure operability
of the main steam line flow circuits and to be consistent with the TSs
surveillance requirements for other engineered safety features
instruments.
These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to
Accommodate a 24-Month Fuel Cycle.''
Date of issuance: January 11, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 142
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41510)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 11, 1994.No significant hazards
consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: October 29, 1993
Brief description of amendment: The amendment revises Technical
Specifications (TSs) Sections 3.10 (Control Rods and Power Distribution
Limits) and 4.2 (Inservice Inspections) to correct administrative
errors that resulted from the issuance of TS Amendment Nos. 57 and 103.
The amendment corrects the errors and further clarifies the TS.
Date of issuance: January 12, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 143
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64615) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 12, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
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 application for amendment: October 30, 1991
Brief description of amendment: The amendment revised Technical
Specification 3.1.3.2, ``Control Rod Maximum Scram Insertion Times,''
to clarify the conditions under which the plant must be shut down in
the event that individual control rod scram insertion times exceed the
allowable values.
Date of issuance: January 19, 1994
Effective date: January 19, 1994
Amendment No. 54
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 11, 1991 (56
FR 64651) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 19, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By March 4, 1994, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC 20555 and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Commonwealth Edison Company, Docket No. 50-265, Quad Cities Nuclear
Power Station, Unit 2, Rock Island County, Illinois
Date of application for amendments: October 29, 1993, as
supplemented by letters dated December 22, 1993 and January 14, 1994.
Brief description of amendments: The license amendment dispositions
Unreviewed Safety Questions (USQ) related to proposed plant
modifications associated with reactor vessel water level
instrumentation. These modifications have been initiated to mitigate
the circumstances outlined in NRC Bulletin 93-03, ``Resolutions of
Issues Related to Reactor Vessel Water Level Instrumentation in BWRs.''
Date of issuance: January 19, 1994
Effective date: January 19, 1994
Amendment No.: 139
Facility Operating License No. DPR-30. Public comments requested as
to proposed no significant hazards consideration: No.The Commission's
related evaluation of the amendment, finding of emergency
circumstances, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated January 19,
1994.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
NRC Project Director: James E. Dyer
Dated at Rockville, Maryland, this 26th day of January 1994.
For the Nuclear Regulatory Commission.
John N. Hannon,
Acting Director, Division of Reactor Projects - III/IV/V, Office of
Nuclear Reactor Regulation
[Doc. 94-2174 Filed 2-1-94; 8:45 am]
BILLING CODE 7590-01-F