[Federal Register Volume 61, Number 206 (Wednesday, October 23, 1996)]
[Notices]
[Pages 55028-55049]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-27025]
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UNITED STATES 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 September 30, 1996, through October 10,
1996. The last biweekly notice was published on October 9, 1996 (61 FR
52962).
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 Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By November 22, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or
[[Page 55029]]
controverted. In addition, the petitioner shall provide a brief
explanation of the bases of the contention and a concise statement of
the alleged facts or expert opinion which support the contention and on
which the petitioner intends to rely in proving the contention at the
hearing. The petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner intends to rely to establish those facts or expert opinion.
Petitioner must provide sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: September 18, 1996
Description of amendment request: Revise Technical Specification
(TS) 4.8.1.1.2 by removing TS 4.8.1.1.2.h.2 pressure testing
requirement since adequate testing will be completed in accordance with
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code, Section XI.
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:
This change does not involve a significant hazards consideration
for the following reasons:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Applying ASME Code, Section XI alternative examination/testing
will not affect any initiators of any previously evaluated accidents
or change the manner in which the emergency diesel generators or any
other systems operate. The diesel fuel oil system supports the
emergency diesel generators which serve an accident mitigating
function. Where portions of piping are non-isolable or where
atmospheric tanks are involved, the Section XI ASME alternatives to
110% pressure testing continue to ensure the integrity of the fuel
oil system without any impact on analyzed accident scenarios or
their consequences. Therefore, the proposed amendment does not
result in an 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.
The proposed alternative testing and surveillance will not
involve any physical alterations or additions to plant equipment or
alter the manner in which any safety-related system performs it
function. Using ASME Section XI, or NRC-approved ASME Code cases, as
guidance for pressure testing continues to provide assurance that
the fuel oil supply system will perform its intended function.
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.
There are no changes being made to the safety limits or safety
settings that would adversely impact plant safety. Further, there is
no impact on the margin of safety as defined in the Technical
Specifications. Utilizing ASME Section XI as guidance for
determining those sections of piping that should be pressure-tested
or tested at atmospheric pressure will ensure proper operation of
the diesel generator fuel oil supply 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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: F. Mark Reinhart, Acting
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power
Plant, Unit 1, Monroe County, Michigan
Date of amendment request: August 29, 1996 (Reference NRC-96-0111)
Description of amendment request: The proposed amendment will: (1)
allow certain equipment and instruments to be removed from service for
short periods of time to allow for
[[Page 55030]]
maintenance, testing, inspection, modifications, and account for
equipment failures; (2) reduce the frequency of environmental liquid
effluent monitoring and eliminate one raw water sampling location; (3)
eliminate the requirement for moisture intrusion monitoring for the
reactor building lower level; and (4) correction of a typographical
error.
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 using the standards in 10 CFR 50.92(c). The licensee's
analysis is presented below:
(1) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident. Provisions for
removing the primary cover gas supply from service for short periods
of time will not significantly increase the probability of an
accident occurring as long as the probability of a significant water
reaction with residual sodium is not significantly increased. This
is ensured by prescribing limits on the time that carbon dioxide
pressure can be low. The consequences of an accident would not be
affected by provisions for removing the primary cover gas supply
from service as this equipment does not mitigate accidents or affect
the accident sequences. Similarly, the provisions for removing the
moisture intrusion and cover gas pressure alarms from service for
short period of time will not significantly increase the probability
of an accident. The alarms provide a monitoring function to detect
degradation in the performance of the cover gas supply and sump
systems. Absence of these alarm functions for short periods of time
does not increase the probability of such degradation and it does
not significantly impact the ability for timely detection of such
degradation. The consequences of an accident would not be affected
by provisions for removing the moisture intrusion and cover gas
pressure alarms from service as this equipment does not mitigate
accidents or affect the accident sequences. Elimination of the
moisture intrusion alarm for the reactor building lower level does
not significantly increase the probability of an accident because
the probability that water could accumulate in this area is
essentially unchanged. Design features of the foundation,
containment structure, and annulus drains are intended to prevent
entry of water into the reactor building. These features have
prevented any water intrusion into this area. The consequences of an
accident would not be affected by elimination of the moisture
intrusion alarm for the reactor building lower level because this
equipment does not mitigate accidents or affect the accident
sequences. The Safety Evaluation Supporting Amendment 9 to the
referenced license did not rely on moisture intrusion monitoring and
alarm features for any safety function or accident prevention or
mitigation function. Environmental monitoring surveillance are
unrelated to postulated accident sequences and cannot affect the
probability or consequences of an accident. The correction of the
typographical error is unrelated to accident initiation and
sequences and cannot affect the probability or consequences of any
accident.
(2) The operation of Enrico Fermi Atomic Power Plant, 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 do not create the possibility of a new or
different accident from any previously evaluated. With the exception
of the allowance for composite environmental samples, which are
unrelated to any potential accident sequence, these changes propose
no new activities or new methods for performing existing activities.
Previous evaluations have considered the release of all of the
radioactivity in the residual sodium due to postulated fire or other
catastrophe and release of radioactive water stored in the liquid
waste tanks which bound the only possible radiological accidents at
Fermi 1. For these reasons, no new or different type of accident is
created by these changes.
(3) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety. The changes to the primary system cover gas system
technical specifications still ensure that any residual sodium is
passivated by carbon dioxide. Changes to the alarms affect only
monitoring functions and therefore do not cause a change to any
parameter that could affect the margin of safety. Similarly, the
environmental surveillances are unrelated to margin of safety. The
correction of the typographical error is unrelated to margin of
safety. For these reasons, the proposed changes do 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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esquire, Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226NRC Branch Chief: Michael F.
Weber
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: September 25, 1996 (NRC-96-0085)
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirement 4.8.4.3 to
remove the requirement to periodically test the thermal overload (TOL)
devices for safety-related motor-operated valves (MOVs). The
surveillance requirement would continue to require testing of a TOL
device following any maintenance activity that could affect the
performance of the device. The surveillance requirement would also be
clarified by indicating that testing of TOL devices is required upon
initial installation. The associated portion of the TS Bases would also
be revised to reflect this change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident. The deletion of
the requirement for testing of the TOL protective devices lessens
degradation to the components which can improve MOV reliability.
Based on historical data through the years of testing, there is no
significant drifting of the trip setpoints of the TOL protective
devices. The probability of an accident would not increase since
terminating the periodic testing or clarifying the situational
testing requirements cannot cause equipment to operate inadvertently
and so cannot cause an accident. The periodic testing of the TOL
protective devices can temporarily render MOVs inoperable due to the
removal of the components from service and can cause safety systems/
divisions to become unavailable. The deletion of the periodic
testing requirement would increase the availability of safety
systems insuring that they would be able to respond to accident
conditions. The consequences of an accident will not increase since
eliminating the periodic testing and clarifying the situational
testing requirements will improve reliability of safety-related MOVs
to respond to an accident and will not increase the failure rate of
equipment. The clarification of the situational testing ensures that
the test will be conducted after any maintenance that could affect
the performance of the TOL protective devices. Thus, the proposed
change increases reliability of the MOVs and increases plant safety.
Therefore this change will not result in a significant increase in
the probability or consequences of an accident.
2. The proposed change does not create the possibility of a new
or different accident from any previously evaluated. The TOL
[[Page 55031]]
protective devices are not an accident initiator, they only protect
equipment provided to mitigate the consequences of an accident. For
this reason, no new or different type of accident is created by this
change.
3. The proposed change does not involve a significant reduction
in a margin of safety. The trip setpoints of the TOL protective
devices depend upon both the current and the length of time the
current is applied. The trip setpoints for TOL protective devices
are much higher than conditions normally experienced during an MOV
stroke and are meant to protect the motor from stall and overload
conditions. The difference between the current of the trip setpoints
and the normal conditions is great enough that a premature trip of
the TOL protective device is highly unlikely, even at degraded
voltages. The TOL protective device protects the motor from the
stall conditions. Not conducting the periodic testing of the TOL
protective devices would not cause the MOVs to fail, nor would the
performance of the MOVs be adversely affected. Throughout the life
of the plant, there has never been an instance of a safety related
MOV failure due to degradation or failure of TOL protective devices.
Further, based on maintenance history, the elimination of the
periodic testing would eliminate any significant potential
degradation of the TOL protective devices, thereby increasing their
reliability. Finally, with the removal of the periodic testing of
the TOL protective devices, fewer MOVs would have to be removed from
service for testing. Since necessary components would no longer be
inoperable due to the periodic testing, there would be an increase
of availability time of safety systems/divisions. Deletion of the
periodic testing could reduce the durations of online system
outages. Clarifying the situational testing requirements would
better define when the testing of the TOL protective devices is
necessary which would ensure operability. The testing would be based
on installation or any maintenance that could affect the TOL
protective device. For these reasons, 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226
NRC Project Director: John N. Hannon
Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear
Station, Units 1 and 2, York County, South Carolina
Date of amendment request: June 21, 1996
Description of amendment request: The proposed amendments would
administratively correct the term ``lifting load'' in Technical
Specification 3.9.6b.2 to ``lifting force.'' This correction would
clarify that the static loads associated with the lifting tool, drive
rod and control rod weights are not included in the lifting force
limit. The amendments would also more accurately define auxiliary hoist
minimum capacities and give a more expansive description of the
activities for which protective measures and surveillance testing are
used.
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:
Question: Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change[s] [are] administrative in nature, and
do[] not represent any changes to the refueling process in the
field. It more accurately describes the components for which the
LCO's [limiting conditions of operation] protection is intended as
well as giving a more accurate description of the auxiliary hoist's
minimum capacity. [They] also broaden[] the domain of activities for
which protective measures are taken, by including drag load testing
into monitored activities. At both MNS [McGuire Nuclear Station] and
CNS [Catawba Nuclear Station], the auxiliary hoists and the
manipulator cranes are rated at [greater than or equal to] 3000
pounds and are surveillance tested to greater than 1000 pounds. This
brackets the limit force lifting value change from 600 to 1000
pounds in the amendment proposal.
Question: Will the change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. Th[ese] proposed administrative change[s] reflect[] no
changes in the refueling processes, or any systems, structures or
components connected with the refueling process.
Question: Will the change involve a significant reduction in a
margin of safety?
No. The proposed administrative change[s] [have] no impact on
refueling processes, systems, structures or components, and do[] not
result in any significant reduction in a margin of safety. The
subject change[s] only clarif[y] the original intent of the
specification and more accurately describe[] the involved
components, component capacities and the domain of activities for
which measures are taken to protect the reactor internals.
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
proposed amendments involve no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: September 17, 1996 (TSC 96-01)
Description of amendment request: The proposed changes would reduce
the Reactor Building pressure setpoint for actuation of the Reactor
Building Spray System in Technical Specification (TS) 3.5.3 from a
maximum of 30 pounds per square inch gauge (psig) to 15 psig, reduce
the maximum allowable Reactor Building internal pressure specified in
TS 3.6.4 from 1.5 psig to 1.2 psig when the reactor is critical, revise
the corresponding Bases of TS 3.3 to indicate that the Reactor Building
sprays and coolers are designed to mitigate the containment temperature
response rather than containment pressure response to a loss-of-coolant
accident, and make other administrative changes. In addition, the lower
Reactor Building pressure limit (a vacuum of 5 inches of mercury (Hg))
in Specification 3.6.4 would be changed to the corresponding value in
terms of psig to reflect the units displayed on the control room
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated:
No. The analysis of the post-LOCA [loss-of-coolant accident]
Reactor Building response to high-energy line breaks, using the new
methodology, uses assumptions different from the requirements
currently delineated in Technical Specifications. The new
assumptions used for initial Reactor Building pressure and Reactor
Building Spray system
[[Page 55032]]
actuation are 1.2 psig and 20 psig respectively. These values are
lower, and hence more conservative, than the values currently
specified in Technical Specifications.
Since the new values for Reactor Building pressure and Reactor
Building Spray actuation are more conservative and the analysis
methodology has received approval from the NRC via [an] SER, this
change does not involve a significant increase in the probability or
consequences of an accident previously identified.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
No. The methodology for Reactor Building high energy line break
analysis is being revised. The revision of the method of analysis
does not alter the manner by which plant systems and components
function for accident mitigation.
(3) Involve a significant reduction in a margin of safety.
No. By letter dated March 15, 1995, the NRC stated that the new
analyses described in the topical report, DPC-NE-3003-P, expand the
scope of analyzed piping failures in containment for the Oconee
facilities. The NRC further stated that this new analysis method has
been used to reanalyze existing licensing basis pipe failure events
in containment, and to examine the potential effects of previously
unanalyzed assumptions and initial conditions which the NRC staff
finds to be consistent with current NRC staff acceptance criteria or
produce equally conservative results. In conclusion, the NRC
confirmed that this methodology, with appropriate adjustments to
reflect potential plant modifications, may be used by Duke Power to
perform future analyses in support of licensing applications related
to containment accident response. This proposed change to Technical
Specifications reflects the use of this new methodology. Based on
this new methodology, changes have been made to setpoint assumptions
for initial Reactor Building pressure and Reactor Building Spray
actuation. This proposed Technical Specification change reflects
those assumption changes. This methodology has been accepted by the
NRC. This proposed change to Technical Specifications does not
involve a significant 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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of amendment request: September 9, 1996
Description of amendment request: The proposed amendment would
revise the Minimum Channels Operable requirement of Item 4.c (Steam
Line Isolation, Containment Pressure Intermediate -- High-High) of
Technical Specification (TS) Table 3.3-3 from 3 to 2. This proposed
change would make this Unit 1 TS consistent with the comparable Unit 2
TS.
The proposed amendment would also revise the minimum charging pump
discharge pressure in TS 3.5.5 from 2311 psig to 2397 psig. This change
is required to ensure that safety analysis assumptions for safety
injection flow are met. Conforming changes would also be made to the
Bases for TS 3/4.5.5 to reflect the proposed changes to TS 3.5.5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment does not add or modify any existing plant
equipment. Since normal charging pump discharge pressure is greater
than or equal to approximately 2440 psig, no additional plant
configuration changes or modifications will be required to comply
with this revised charging pump discharge pressure value. The
proposed amendment does not change the design or function of the
containment pressure intermediate-high-high channels.
The consequences of an accident previously evaluated are not
significantly increased. The ability of the containment pressure
intermediate-high-high function to initiate steam line isolation
will not be affected. Since steam line isolation will continue to
occur at the same required trip setpoint, the amount of mass and
energy released to containment along with the ability to maintain at
least one unfaulted steam generator (SG) as a heat sink for the
reactor remains unchanged. The amount of seal injection flow will
continue to be adequately limited to ensure sufficient flow to the
reactor core during accident conditions. The Bases changes are
editorial in nature and do not involve a change to probability or
consequences of an accident previously evaluated.
Based on the above discussion, it is concluded that this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed amendment does not change the plant configuration
in a way which introduces a new potential hazard to the plant. Since
design requirements continue to be met and the integrity of the
reactor coolant system pressure boundary is not challenged, no new
failure mode has been created. As a result, an accident which is
different than already evaluated in the Updated Final Safety
Analysis Report will not be created due to this change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety is not significantly reduced by this
proposed change. The trip setpoint for the containment pressure
intermediate-high-high function remains unchanged. With one channel
inoperable, the remaining two channels will continue to initiate the
protective function on a two-out-of-two logic. The action statement
limits this condition to 6 hours after which time the inoperable
channel must be placed in the trip condition. This action restores
the function to be able to meet single failure criteria on a one-
out-of-two logic basis.
The proposed revision to the charging pump discharge pressure
will not change the flow limit on seal injection. The specification
will continue to ensure that seal injection flow is limited. This
will ensure that sufficient flow to the reactor core is provided
during accident conditions.
The proposed changes to the Bases for seal injection flow are
editorial in nature and do not affect the margin of safety.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Entergy Gulf States Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 29, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) to reflect the elimination of
T-factor adjustments in the Average Power
[[Page 55033]]
Range Monitors (APRM) setpoints, a decrease in the calibration
frequency of the Local Power Range Monitors (LPMR), and an improvement
in the calculation of Reactivity Anomaly.
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 request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
This change replaces the APRM setpoints T-factor limit with
power and flow-dependent minimum critical power ratio (MCPR) and
linear heat generation rate (LHGR) limits. These new power and flow-
dependent thermal limits eliminate the need for manual setpoint
adjustment resulting from power peaking conditions. The new power
and flow-dependent thermal limits are automatically applied by
computer software during the calculation of the core thermal limits
and, therefore, do not require manual setpoint adjustments based on
the power peaking conditions in the reactor. Extensive transient
analyses at a variety of power and flow conditions have been
performed and were utilized to study the trend of transient severity
without the setpoints T-factor limit. A large data base was
established by analyzing limiting transients over a range of power
and flow conditions. The data base included evaluations
representative of a variety of plant configurations and parameters
such that the conclusions drawn from the studies would be applicable
to the broad range of boiling water reactors (BWRs). This data base
was utilized to develop plant specific operating limits (MCPR and
LHGR), which assures that margins to fuel safety limits are equal to
or larger than those currently in existence with the APRM setpoints
T-factor limit applied. Therefore, this change does not involve an
increase in the probability of any event previously evaluated.
The consequences of an accident previously evaluated have not
been increased because, in all cases, the new power and flow-
dependent thermal limits (MCPR and LHGR) assure that margins to fuel
safety limits are equal to or larger than those currently in
existence with the APRM setpoints T-factor limit applied. Protection
of other thermal limits for all previously analyzed events is
accomplished by specific limits that are independent of the APRM
setpoints T-factor. These are the power and flow-dependent MCPR
Operating Limits which provide protection from fuel dryout and the
rated maximum average planner linear heat generation rate (MAPLHGR)
limit which provides protection of the peak clad temperature for the
design basis accident-loss of coolant accident (DBA LOCA).
Therefore, the proposed change does not involve a significant
increase in the consequences of any event previously evaluated.
No new equipment is introduced by the change in the local power
range monitor (LPRM) calibration frequency and, therefore, the
probability for an accident previously evaluated is unchanged. The
consequences of an accident can be affected by the thermal limits
prior to the accident but LPRM chamber and cycle exposure have no
significant effect on the calculated thermal limits. The thermal
limit calculation is not significantly effected because the LPRM
sensitivity versus exposure function is well defined. This allows
accurate LPRM end-of-life calculations so that detectors can be
replaced before their behavior significantly deteriorates. In the
event deterioration is noted late in the cycle for a few chambers,
they can be bypassed with no significant effect on uncertainties.
Also, the total nodal power uncertainty remains less than the
uncertainty assumed in the General Electric BWR Thermal Analysis
Basis (GETAB) safety limit. Therefore, the thermal limit calculation
is not affected by the LPRM calibration frequency and the
consequences of an accident previously evaluated are not changed.
The change in the parameters used to measure reactivity for
calculation of the reactivity anomaly has no affect on either the
consequences or the probability of an accident previously evaluated
because the allowed reactivity anomaly criteria is unchanged. The
only change is the parameters used to measure reactivity.
Therefore, the proposed elimination of the APRM setpoints T-
factor maintains adequate off-rated MCPR and LHGR margin for all
operating conditions. Also, the change in the LPRM calibration
frequency continues to maintain the accuracy of the thermal limit
calculation. Therefore, the consequences of an accident previously
evaluated are not affected by this change. Finally, the change in
the parameters used to measure reactivity for calculation of the
reactivity anomaly has no affect on either the consequences nor the
probability of an accident previously evaluated. Since no new plant
equipment is introduced by any of the proposed changes, the
probability of accidents previously evaluated are not changed.
Therefore, none of the proposed changes involve an increase in the
probability or consequences of any event previously evaluated.
2. The request does not create the possibility of occurrence of
a new or different kind of accident from any accident previously
evaluated.
This change only replaces the APRM setpoints T-factor limit with
power and flow-dependent MCPR and LHGR limits, changes the LPRM
calibration frequency, and a change to the parameter(s) used to
measure reactivity. None of the proposed changes involve any new
modes of operation or any plant modifications. Therefore, the
proposed changes do not create the possibility of a new or different
type of accident from any accident previously analyzed.
3. The request does not involve a significant reduction in a
margin of safety.
The replacement of the APRM setpoints T-factor limit with power
and flow-dependent thermal limits has been confirmed to provide
adequate MCPR and LHGR protection at all reactor operation
conditions. Operation with higher peaking without APRM gains or flow
bias trip setpoints adjustment does not involve a reduction in a
margin of safety because the higher power peaking resulting from
elimination of the APRM setpoints T-factor has been analyzed to
assure that the margins to fuel safety limits are equal to or larger
than those currently in existence with the APRM setpoints T-factor
limit applied. Therefore, the replacement of the APRM setpoint T-
factor with power and flow-dependent thermal limits does not involve
a reduction in the margin of safety.
Protection of other thermal limits for all previously analyzed
events is accomplished by specific limits that are independent of
the APRM setpoint T-factor limit. These are the power and flow-
dependent
MCPR Operating Limits which provide protection from fuel dryout
and the rated MAPLHGR limit which provides protection of the peak
clad temperature for the DBA LOCA.
The margin of safety can be affected by the thermal limits prior
to an accident but LPRM chamber exposure and cycle exposure have no
significant effect on the calculated thermal limits. The thermal
limit calculation is not significantly affected because the LPRM
sensitivity versus exposure function is well defined. This allows
accurate LPRM end of life calculations so that detectors can be
replaced before their behavior significantly deteriorates. In the
event deterioration is noted late in the cycle for a few chambers,
they can be bypassed with no significant effect on uncertainties.
Also, the total nodal power uncertainty remains less than the
uncertainty assumed in the GETAB safety limit. Therefore neither the
thermal limit calculation nor the margin of safety are affected by
the LPRM calibration.
The change in the parameters used to measure reactivity for
calculation of the reactivity anomaly has no affect on the margin of
safety because the allowed reactivity anomaly criteria is unchanged.
The only change is the parameters used to measure reactivity.
Neither the change to APRM setpoints T-factor nor the change to
the LPRM calibration frequency significantly effects the thermal
limits calculation, and, therefore, do not result in an increase in
core damage frequency. The change in the parameters used to measure
reactivity for calculation of the reactivity anomaly has no affect
on the core damage frequency because the allowable reactivity
anomaly criteria remains unchanged. Therefore, the proposed changes
do not involve a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Documenmt Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
[[Page 55034]]
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 29, 1996
Description of amendment request: The proposed amendment would
provide a revision to the reactor pressure vessel (RPV) surveillance
capsule withdrawal schedule for the River Bend Station. The first
surveillance capsule would be withdrawn at 10.4 effective full power
years (EFPY) rather than at 6EFPY.
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.
Pressure-temperature (P-T) limits (RBS Technical Specifications
Figure 3.4.11-1) are imposed on the reactor coolant system to ensure
that adequate safety margins against nonductile or rapidly
propagating failure exist during normal operation, anticipated
operational occurrences, and system hydrostatic tests. The P-T
limits are related to the nil-ductility reference temperature,
RTNDT, as described in ASME Section III, Appendix G. Changes in
the fracture toughness properties of RPV beltline materials,
resulting from the neutron irradiation and the thermal environment,
are monitored by a surveillance program in compliance with the
requirements of 10CFR50, Appendix H. The effect of neutron fluence
on the shift in the nil-ductility reference temperature of pressure
vessel steel is predicted by methods give in Regulatory Guide 1.99,
Rev. 2.
River Bend's current P-T limits were established based on
adjusted reference temperatures developed in accordance with the
procedures prescribed in Reg. Guide 1.99, Rev. 2, Regulatory
Position 1. Calculation of adjusted reference temperature by these
procedures includes a margin term to ensure conservative, upper-
bound values are used for the calculation of the P-T limits.
Revision of the first capsule withdrawal schedule will not affect
the P-T limits because they will continue to be established in
accordance with Regulatory Position 1 (or other NRC-approved)
procedures. When permitted (two or more credible surveillance data
sets available), Regulatory Position 2 (or other NRC-approved)
methods for determining adjusted reference temperature will be
followed.
This change is not related to any accidents previously
evaluated. The proposed change is a revision of the Withdrawal Time
for the first surveillance capsule as given in Technical
Requirements (TR) Table 3.4.11-1 from 6 EFPY to 10.4 EFPY. This
change will not affect P-T limits as given in RBS Technical
Specifications Figure 3.4.11-1 or USAR Figures 5.3-4a and 5.3-4b.
This change will not affect any plant safety limits or limiting
conditions of operation. The proposed change will not affect reactor
pressure vessel performance as no physical changes are involved and
RBS vessel P-T limits will remain conservative in accordance with
Reg. Guide 1.99, Rev. 2 requirements. The proposed change will not
cause the reactor pressure vessel or interfacing systems to be
operated outside of their design or testing limits. Also, the
proposed change will not alter any assumptions previously made in
evaluating the radiological consequences of accidents. Therefore,
the probability or consequences of accidents previously evaluated
will not be increased by the proposed change.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change is a revision of the Withdrawal Time in TR
Table 3.4.11 for the first RPV material surveillance capsule from 6
EFPY to 10.4 EFPY. This proposed change does not involve a
modification of the design of plant structures, systems, or
components. The proposed change will not impact the manner in which
the plant is operated as plant operating and testing procedures will
not be affected by the change. The proposed change will not degrade
the reliability of structures, systems or components important to
safety (ITS) as equipment protection features will not be deleted or
modified, equipment redundancy or independence will not be reduced,
supporting system performance will not be downgraded, the frequency
of operation of ITS equipment will not be increased, and increased
or more severe testing of ITS equipment will not be imposed. No new
accident types or failure modes will be introduced as a result of
the proposed change. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from that
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
As stated in the River Bend SER, ``Appendices G and H of 10CFR50
describe the conditions that require pressure-temperature limits and
provide the general bases for these limits. These appendices
specifically require that pressure-temperature limits must provide
safety margins at least as great as those recommended in the ASME
Code, Section III, Appendix G. .... Until the results from the
reactor vessel surveillance program become available, the staff will
use RG 1.99, Revision 1 [now Revision 2] to predict the amount of
neutron irradiation damage. ... The use of operating limits based on
these criteria--as defined by applicable regulations, codes, and
standards--will provide reasonable assurance that nonductile or
rapidly propagating failure will not occur, and will constitute an
acceptable basis for satisfying the applicable requirements of GDC
31.''
Bases for RBS Technical Specification 3/4/11 states: ``The P/T
limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering
pressure, temperature, and temperature rate of change conditions
that might cause undetected flaws to propagate and cause nonductile
failure of the RCPB [Reactor Coolant Pressure Boundary], a condition
that is unanalyzed. ... Since the P/T limits are not derived from
any DBA, there are no acceptance limits related to the P/T limits.
Rather, the P/T limits are acceptance limits themselves since they
preclude operation in an unanalyzed condition.''
The proposed change will not affect any safety limits, limiting
safety system settings, or limiting conditions of operation. The
proposed change does not represent a change in initial conditions,
or in a system response time, or in any other parameter affecting
the course of an accident analysis supporting the Bases of any
Technical Specification. The proposed change does not involve
revision of the P-T limits but rather a revision of the Withdrawal
Time for the first surveillance capsule. The current P-T limits were
established based on adjusted reference temperatures for vessel
beltline materials calculated in accordance with Regulatory Position
1 of Reg. Guide 1.99, Rev. 2. P-T limits will continue to be revised
as necessary for changes in adjusted reference temperature due to
changes in fluence according to Regulatory Position 1 until two or
more credible surveillance data sets become available. When two or
more credible surveillance data sets become available, P-T limits
will be revised as prescribed by Regulatory Position 2 of Reg. Guide
1.99, Rev. 2 or other NRC-approved guidance. Therefore, the proposed
changes do not involve a significant reduction in any margins of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: September 23, 1996
Description of amendment request: The proposed amendment would
revise the Crystal River Unit 3 (CR 3) technical specifications (TS) to
delete a note
[[Page 55035]]
associated with Surveillance Requirement (SR) 3.3.7.1 for the
Engineered Safeguard Actuation System (ESAS) Automatic Actuation Logic.
Applicable TS Bases will also be revised to reflect the proposed TS
change.
SR 3.3.7.1 requires periodic testing of the ESAS automatic
actuation logic matrix to demonstrate that the required logic
combinations are operable. When the ESAS automatic actuation logic is
placed in an inoperable status solely for performing of this
surveillance, the note associated with the SR 3.3.7.1 provides relief
in that it allows not entering into applicable Conditions and Required
Actions for up to 8 hours, provided the associated engineering
safeguards (ES) function is maintained. The licensee has determined
that because of the CR 3 design of the ESAS System and the way the test
is performed, maintenance of the ``associated ES function'' is not
possible. Thus, the note does not provide the relief intended and
therefore, the licensee proposes to delete the note. During the
performance of the ESAS test and bypassing the associated ES function,
the licensee proposes to enter into applicable TS 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, which is presented below:
1. The proposed change will not significantly increase the
probability or consequences of an accident previously evaluated
because unavailability of equipment is recognized in the design of
the plant and in the Technical Specifications. The probability and
consequences of accidents previously evaluated are bounded by the
evaluations done for the allowed outage time of the associated
functions.
2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously evaluated
because the bypassing of ES functions for testing purposes does not
place the plant in a configuration which would allow the possibility
of a new or different kind or accident to be created.
3. The proposed change will not involve a significant reduction
to the margin of safety because deleting the NOTE does not effect
the way the test is performed. The test is required by the Technical
Specifications and will still be performed in the same manner. Thus,
there is no change in the unavailability of the system as a result
of this change and the margin of safety is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Attorney for licensee: A. H. Stephens, General Counsel, Florida
Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida
33733
NRC Project Director: Frederick J. Hebdon
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: September 27, 1996
Description of amendment request: The proposed amendment would
revise the Crystal River 3 (CR3) post-accident monitoring (PAM)
instrumentation technical specification (TS). Specifically, the
following TS changes are proposed:
A. Table 3.3.17-1, Function 8: The descriptor is changed from
``Containment Pressure (Narrow Range)'' to ``Containment Pressure
(Expected Post-Accident Range).''
B. Table 3.3.17-1, Function 18: The required channels for Core Exit
Temperature (Backup) is changed from ``2 sets of 5'' to ``3 per core
quadrant.''
C. Table 3.3.17-1: A new Function 20 is added and designated as
``Low Pressure Injection Flow.''
D. Table 3.3.17-1: A new Function 21 is added and designated as
``Degrees of Subcooling.''
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 letters A, B, C and D correspond to the proposed TS
changes), which is presented below:
1. The proposed changes will not significantly increase the
probability or consequences of an accident previously evaluated
because:
A/B. The changes in containment pressure and core exit
thermocouple nomenclature do not reflect any physical changes to the
facility.
C/D.The addition of low pressure injection flow and degrees of
subcooling to the Post-Accident Monitoring Instrumentation LCO is
being done to comply with a commitment made during the technical
specification improvement program to include in the technical
specifications, that instrumentation which monitors variables
classified as Type A in accordance with Regulatory Guide 1.97. These
two variables have recently been re-classified as Type A. The
associated instruments are used after an accident occurs to prompt
the operators to take certain mitigative actions. Therefore, the
probability of an accident occurring is unaffected. As part of the
re-classification of these variables to Type A, the associated
monitoring instrumentation will be under more strict surveillance
and control, which provides additional assurance that the prescribed
manual operator actions will be implemented when necessary. This, in
turn, assures the previously evaluated accident consequences remain
valid.
2. The proposed changes will not create the possibility of a new
or different kind of accident from any accident previously evaluated
because:
A/B. The changes in containment pressure and core exit
thermocouple nomenclature do not reflect any physical changes to the
facility. The changes provide clarification for the instruments
which are required to comply with the LCO.
C/D. The addition of low pressure injection flow and degrees of
subcooling to the Post-Accident Monitoring instrumentation LCO is
being done to comply with a commitment made during the technical
specification improvement program to include in the technical
specifications, that instrumentation which monitors variables
classified as Type A in accordance with Regulatory Guide 1.97. These
two variables have been re-classified as Type A. The associated
instruments are used after an accident occurs to prompt the
operators to take certain mitigative actions. Since the
instrumentation is used only post-accident, these changes do not
create the possibility of a new or different kind of accident.
3. The proposed change will not involve a significant reduction
to the margin of safety because:
A/B. The changes in containment pressure and core exit
thermocouple nomenclature have no affect on the margin of safety.
The changes provide clarification of the technical specifications.
This reduces the potential for confusion regarding this
instrumentation.
C/D. The addition of low pressure injection flow and degrees of
subcooling to the post-accident monitoring instrumentation table
adds controls on the OPERABILITY of post-accident monitoring
instrumentation providing greater assurance it will be available
should an accident occur.
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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Attorney for licensee: A. H. Stephens, General Counsel, Florida
Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida
33733
NRC Project Director: Frederick J. Hebdon
[[Page 55036]]
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: September 5, 1996
Description of amendment request: The proposed change deletes
License Condition 2.C.5, Integrated Implementation Schedule.
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 10CFR50.92, NNECO has reviewed the attached
proposed change and has concluded that it does not involve a
significant hazards consideration (SHC). The basis for this is that
the three criteria of 10CFR50.92(c) are not compromised. The
proposed change does not involve an SHC because the change would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Operation of the facility in accordance with the proposed change
would result in a change in an administrative process for
prioritizing and scheduling projects and engineering evaluations.
With the limited number of NRC required projects remaining to be
implemented, the IIS [Integrated Implementation Schedule] is no
longer required to schedule resources for the remaining topics.
Since this license condition only involves an administrative
process, it does not directly affect the design or operation of the
plant. Therefore, no accident analyses are affected by the change,
and the change does not increase the probability or consequences of
any previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed license modification removes a requirement relating
to the scheduling of modifications and engineering evaluations.
Because the license condition addresses only an administrative
scheduling mechanism, it does not affect directly the design or
operation of the plant. Therefore, the proposed change does not
create a different kind of accident from those previously analyzed.
3. Involve a significant reduction in a margin of safety.
The proposed license modification removes a requirement relating
to the scheduling of modifications and engineering evaluations. The
original purpose of the IIS and the ISAP [Integrated Safety
Assessment Program] was to prioritize and schedule modifications and
engineering evaluations in a manner that was agreed upon by both
NNECO and the NRC. These programs were especially important to
Millstone Unit No. 1 for priorization of topics associated with the
SEP [Systematic Evaluation Program] and the TMI [Three Mile Island]
Action Plan. This program is considered to be no longer necessary.
Modifications and engineering evaluations will be scheduled and
prioritized using other methodologies. Since this change involves an
administrative process only, there is no direct impact on the design
or operation of the plant, and therefore, no significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: August 27, 1996
Description of amendment request: The proposed amendment revises
the required value of control rod drive (CRD) system pressure in
technical specification (TS) 3.10.8, ``Shutdown Margin (SDM) Test-
Refueling.''
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 any accident previously
evaluated.
The proposed changes are purely administrative and do not
involve any physical changes to plant SSC [systems, structures and
components]. The change in the minimum CRD charging water header
pressure from 955 psig to 940 psig was previously approved in TS
Amendments Nos. 211 and 216 for PBAPS [Peach Bottom Atomic Power
Station], Units 2 and 3. TS Change Request 95-12 was incomplete by
inadvertently failing to identify the need to change requirement (f)
of LCO [Limiting Condition for Operation] 3.10.8. Therefore, the
proposed changes will not increase the probability of occurrence or
the consequences of an accident previously evaluated in the SAR
[safety analysis report].
2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are purely administrative and do not
involve any physical changes to plant SSC. The proposed changes do
not allow plant operation in any mode that is not already evaluated
in the SAR. Therefore, the possibility of a different type of
accident than previously evaluated in the SAR is not created.
3) The proposed changes do not result in a significant reduction
in the margin of safety.
The proposed changes are purely administrative and have no
impact on any safety analysis assumptions or margins of safety. A
change to SR 3.10.8.6 was approved by the NRC by TS Amendment Nos.
211 and 216. LCO 3.10.8 requirement (f) should have been changed at
the same time to reflect a minimum CRD charging water pressure of
940 psig. Changing LCO 3.10.8 requirement (f) to reflect TS
Amendment Nos. 211 and 216 is purely administrative, and therefore,
does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: May 20, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Sections 3/4.4.9.2, 3/
4.9.11.1, 3/4.9.11.2, and the associated TS Bases 3/4.4.9 and 3/4.9.11,
to more clearly describe that the Residual Heat Removal (RHR) system
Shutdown Cooling mode of operation consists of four (4) ``subsystems.''
These TS sections pertain to plant operations during Operational
Conditions (OPCONs) 4, ``Cold Shutdown'' and 5, ``Refueling.'' In
addition, the proposed TS change would make administrative changes to
TS Section 3/4.4.9.1 to
[[Page 55037]]
ensure consistency in terminology regarding the description of Shutdown
Cooling ``subsystems.'' The proposed TS changes are consistent with the
guidance delineated in the Improved TS (i.e., NUREG-1433, Revision 1,
``Standard Technical Specifications General Electric Plants, BWR/4,''
dated April 1995) which indicates that the RHR Shutdown Cooling mode of
operation is comprised of two (2) loops and four (4) subsystems (i.e.,
two (2) subsystems per loop).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes do not involve any physical changes to
plant structures systems, or components. The RHR [Residual Heat
Removal] Shutdown Cooling mode of operation is manually controlled
and is not required for accident mitigation. The RHR system will
continue to function as designed in all modes of operation. The
consequences of equipment malfunction are not changed from those in
existing analyses, with no increase in onsite or offsite
radiological effects. The RHR system will continue to function as
designed to mitigate the consequences of an accident and resultant
onsite and offsite radiological effects remain as previously
evaluated. The proposed TS changes will revise the TS to more
clearly describe the RHR system configuration in OPCONs 4 and 5. The
proposed changes are consistent with the guidance stipulated in
NUREG-1433, Revision 1.
The four (4) ``subsystem'' Shutdown Cooling designation permits
operability of only one (1) RHR heat exchanger for Shutdown Cooling
service in Operational Conditions (OPCONs) 4 and 5, as long as both
associated RHR pumps are operable and alignable for Shutdown
Cooling. TS requirements for RHR Shutdown Cooling operation in Hot
Shutdown, Suppression Pool Spray, and Suppression Pool Cooling
continue to require two (2) independent loops to be operable in
OPCONs 1, 2, and 3*, meaning both RHR heat exchangers will still be
required to be operable throughout OPCON 3.
The four (4) ``subsystem'' Shutdown Cooling designation has no
effect on the required operability of the Residual Heat Removal
Service Water (RHRSW) system. As required by TS Section 3.7.1.1, the
RHRSW subsystem(s) associated with the required operable RHR heat
exchanger(s) will continue to remain operable. Each operable RHRSW
subsystem consists of two (2) operable pumps and the required
operable flowpath to provide decay heat removal via the associated
RHR heat exchanger.
The RHRSW system piping is designed, fabricated, inspected, and
tested in accordance with the requirements of ASME [American Society
of Mechanical Engineers], Section III Class 3, and each RHRSW
subsystem is single active failure proof in that the failure of a
motor-operated valve, diesel generator, or pump does not prevent the
system from performing its safety function.
The required availability of four (4) loops of the Low Pressure
Coolant Injection (LPCI) mode of RHR during OPCONs 1, 2, and 3 as
required by TS Section 3.5.1 is not impacted by the four (4)
``subsystem'' Shutdown Cooling designation. No change to any RHR
system instrumentation logic, required Emergency Core Cooling System
(ECCS) availability, or method of operation is involved.
NUREG-1433, Revision 1, also re-affirms that each Shutdown
Cooling ``subsystem'' is considered operable if it can be manually
aligned, remotely or locally, in the shutdown cooling mode for
removal of decay heat. Thus, a LPCI-dedicated pump can be aligned
for LPCI automatic initiation, yet still be considered part of an
operable shutdown cooling subsystem as long as it can be re-aligned
for Shutdown Cooling.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes do not involve any physical changes to
plant structures, systems, or components. The RHR system will
continue to function as designed in all modes of operation. No new
accident type is created as a result of the proposed changes. No new
failure mode for any equipment is created. The changes are
consistent with the guidance provided in NUREG-1433, Revision 1,
pertaining to RHR Shutdown Cooling operation in OPCONs 4 and 5.
The four (4) ``subsystem'' Shutdown Cooling designation has no
effect on the required operability of the RHRSW system. The RHRSW
subsystem(s) associated with the required operable RHR heat
exchanger(s) will continue to remain operable as required by TS
Section 3.7.1.1. Each operable RHRSW subsystem consists of two (2)
operable pumps and the required operable flowpath to provide decay
heat removal via the associated RHR heat exchanger.
The RHRSW system piping is designed, fabricated, inspected, and
tested in accordance with the requirements of ASME, Section III,
Class 3, and each RHRSW subsystem is single active failure proof in
that the failure of a motor-operated valve, diesel generator, or
pump does not prevent the system from performing its safety
function.
The required availability of four (4) loops of the LPCI mode of
RHR during OPCONs 1, 2, and 3 as required by TS Section 3.5.1 and
3.5.2 is not impacted by the four (4) ``subsystem'' Shutdown Cooling
designation. No change to any RHR system instrumentation logic,
required ECCS availability, or method of operation is involved.
NUREG-1433, Revision 1, also re-affirms that each Shutdown
Cooling ``subsystem'' is considered operable if it can be manually
aligned, remotely or locally, in the Shutdown Cooling mode for
removal of decay heat. Thus, a LPCI-dedicated pump can aligned be
[sic] [be aligned] for automatic LPCI initiation, yet still be
considered part of an operable shutdown cooling subsystem as long as
it can be re-aligned for Shutdown Cooling.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
Although the Bases for TS Sections 3/4.4.9.2, 3/4.9.11.1, and 3/
4.9.11.2 are being revised in support of this proposed TS change,
the changes only involve providing clarification regarding the
designation of the RHR Shutdown Cooling operation configuration in
OPCONs 4 and 5. The proposed TS changes do not involve any physical
changes to plant structures, systems, or components. The RHR system
will continue to function as designed in all modes of operation. The
consequences of equipment malfunction are not changed from those in
existing analyses, with no increase in onsite or offsite
radiological effects. The RHR system will continue to function as
designed to mitigate the consequences of an accident and resultant
onsite and offsite radiological effects remain as previously
evaluated. The proposed changes are consistent with the guidance
stipulated in NUREG-1433, Revision 1.
The four (4) ``subsystem'' Shutdown Cooling designation has no
effect on the required operability of the RHRSW system. As required
by TS 3.7.1.1, the RHRSW subsystem(s) associated with the required
operable RHR heat exchanger(s) will continue to remain operable.
Each operable RHRSW subsystem consists of two (2) operable pumps and
the required operable flowpath to provide decay heat removal via the
associated RHR heat exchanger.
The RHRSW system piping is designed, fabricated, inspected, and
tested in accordance with the requirements of ASME, Section III,
Class 3, and each RHRSW subsystem is single active failure proof in
that the failure of a motor-operated valve, diesel generator, or
pump does not prevent the system from performing its safety
function. (In the same manner that manual action may be required for
RHR system alignment in OPCONs 4 and 5 with one (1) RHR heat
exchanger operable, a failure of the motor-operated RHRSW inlet or
outlet heat exchanger isolation valves may require manual
positioning for the required alignment.)
The required availability of four (4) loops of the LPCI mode of
RHR during OPCONs 1, 2, and 3* as required by TS Section 3.5.1 is
not affected by the four (4) ``subsystem'' Shutdown Cooling
configuration. No change to any RHR system instrumentation logic,
required ECCS availability, or method of operation is involved.
NUREG-1433, Revision 1, also re-affirms that each Shutdown
Cooling ``subsystem'' is
[[Page 55038]]
considered operable if it can be manually aligned, remotely or
locally, in the Shutdown Cooling mode for removal of decay heat.
Thus, a LPCI-dedicated pump can be aligned for LPCI automatic
initiation, yet still be considered part of an operable Shutdown
Cooling ``subsystem'' as long as it can be re-aligned for Shutdown
Cooling.
Therefore, the proposed TS 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: June 28, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would incorporate performance-based
testing, in accordance with 10 CFR Part 50, Appendix J, ``Primary
Reactor Containment Leakage Testing For Water-Cooled Power Reactors,''
Option B. This option allows utilities to extend the frequencies of the
Type A Containment (ILRT) Leak Rate Test and Type B and C Local Leak
Rate Tests (LLRTs) based on the performance and design of the
containment and components.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Incorporation of the new 10 CFR 50, Appendix J, Option B at LGS,
Units 1 and 2 does not increase the probability of occurrence of an
accident previously evaluated. The containment structure including
its isolation capability is not an accident initiator.
These changes do not involve any changes to the containment
structure, system or components which could increase the probability
of occurrence of an accident previously evaluated or act as a new
accident initiator. Implementation of the proposed changes will
affect the manner in which these structures, systems, or components
(SSCs) are tested; however, the new testing schedule is not an
initiator of any analyzed event. No equipment changes are involved
with adoption of Option B; therefore, performance-based test
intervals for Type A, B, and C tests do not increase the probability
of occurrence of a malfunction of equipment important to safety
previously evaluated. No physical changes are being made to the
plant, nor are there any changes being made in the operation of the
plant as the result of increasing the test intervals. Additionally,
the proposed TS changes will not alter the operation of equipment
available for the mitigation of accidents or transients, therefore,
this change will not result in any significant increase to onsite or
offsite dose previously evaluated. The potential for time-based and
activity-based failure mechanisms which could lead to excessive
containment leakage has been determined to be minimal. Performance-
based test intervals for Type A, B, and C tests will not alter any
safety limits which ensure the integrity of fuel barriers, and will
not increase the primary containment leakage limits.
Performance-based test intervals for Type A, B, and C leak tests
do not increase the consequences of an accident previously
evaluated. NUREG-1493 concluded that reducing the frequency of Type
A tests from the current three per ten years to one per ten years
was found to lead to an imperceptible increase in risk. NUREG-1493
includes the results of a sensitivity study performed to explore the
risk impact of several alternative leak rate test schedules. The
estimated increase in population exposure risk ranged from 0.02% to
0.14%. The risk impact was determined to be very small since Type B
and C testing (local leak rate tests) detect a very large percentage
of overall containment leakages. The percentage of leakages detected
by Type A tests is very small. Past test results experienced at
Limerick Units 1 and 2 concur with these determinations. NUREG-1493
also concluded that the overall unit risk is not very sensitive to
changes in containment leakage rates. Given the insensitivity of
risk to containment leak rates and the small fraction of leak paths
detected solely by the Type A tests, increasing the interval between
Type A tests is possible with minimal impact on public risk.
NUREG-1493 also concluded that, based on a model of component
failure with time, the performance-based alternatives to current,
local-leakage testing requirements are feasible without significant
risk impact. The LGS design and past performance is bounded by the
NUREG study. The NUREG model indicated that the number of components
tested could be reduced by about 60% with less than a three-fold
increase in the incremental risk due to containment leakage. Since
under existing requirements, leakage contributes less than 0.1
percent of overall accident risk, the overall impact is very small.
Therefore, the proposed TS changes will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Performance-based test intervals for Type A, B, and C leak tests
do not introduce a new or different type of accident or create the
possibility of a different type of malfunction of equipment
important to safety than previously evaluated. No physical changes
are being made to the plant, nor are there any changes being made in
the operation of the plant as the result of increasing the test
intervals. No new failure modes of plant equipment previously
evaluated will be introduced. Additionally, the TS changes will not
alter the operation of equipment available for the mitigation of
accidents or transients. The safety function of the primary
containment will be retained since the containment will continue to
provide an essentially leak tight barrier against the uncontrolled
release of radioactivity to the environment for postulated accidents
previously evaluated.
Therefore, the proposed TS changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety is not reduced as a result of adopting 10
CFR 50, Appendix J, Option B. The effect of increasing containment
leakage rate testing intervals was evaluated in NUREG-1493 using
historical industry leakage rate testing results. Performance
history at LGS is consistent with the conclusions reached in NUREG-
1493 and NEI 94-01. The results of the NUREG evaluation conclude
that the increased safety risk corresponding to the extended test
intervals is small (less than 0.1% of total risk). The revised TS
will continue to maintain the allowable leakage rate for the Type A
tests. In addition, the requirement to perform a periodic general
visual inspection of the primary containment has been maintained at
the original interval of three times in 10 years as part of the
performance-based leakage rate testing program.
The risk of a non-detectable increase of primary containment
leakage is considered to be negligible due to the conclusion that 10
CFR 50, Appendix J, Type B and C testing program will continue to be
conducted between Type A tests. A review of previous LGS Type A test
results has concluded that the only failure mechanisms are activity-
based. There is no indication of time-based failures that would not
be identified during the performance of Type B and C tests.
Therefore, we have concluded that the proposed adoption of the
Option B intervals would not result in a non-detectable primary
containment leakage rate in excess of the allowable value (i.e.,
0.5% wt/day) established by the LGS TS.
The proposed TS will continue to maintain the allowable leakage
rate for the combined Type B and C tests. As supported by the
findings of NUREG-1493, the percentage of leakages detected by Type
A tests is small (as
[[Page 55039]]
stated above) and Type B and C leakage tests are capable of
detecting more than 97% of containment leakages and virtually all
such leakages are identified by local leak rate tests of containment
isolation valves. The Type B and C test intervals will be
established through the PCLRTP for each component based on design
and previous LGS test performance history.
Therefore, the proposed TS changes do not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: September 25, 1996
Description of amendment request: The amendments would relocate to
the Salem Updated Final Safety Analysis Report the list of containment
isolation valves that are currently located in Table 3.6-1 of Technical
Specification 3.6.3. In addition, references to the table in
specifications 1.7, 3.6.1, and 3.6.3 are being updated.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequence of an accident previously
evaluated.
The proposed changes simplify the TS, meet the regulatory
requirements for control of containment isolation, and are
consistent with the guidance provided in Generic Letter (GL) 91-08,
``Removal of Component Lists from Technical Specifications.'' The
procedural details of TS Table 3.6-1 have not been changed, only
relocated to a different controlling document, the Salem Update
[sic] [Updated] Final Safety Analysis Report (UFSAR). The proposed
changes are administrative in nature, should result in improved
administrative practices, and do not affect plant operations.
The probability of occurrence of a previously evaluated accident
is not increased because this change does not introduce any new
potential accident initiating conditions. The consequences of an
accident previously evaluated is not increased because the ability
of containment to restrict the release of any fission product
radioactivity to the environment will not be degraded by this
change.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, do not result
in a physical alterations or changes to the operation of the plant,
and cause no change in the method by which any safety-related system
performs its functions. Therefore, this proposed change will not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The administrative change to relocate TS Table 3.6-1 to the
UFSAR does not alter the basic regulatory requirements for
containment isolation and will not adversely affect the containment
isolation capability for credible accident scenarios. Adequate
control of the content of the relocated table is assured by the
10CFR50.59 review process.
The proposed relocation of TS Table 3.6-1 does not alter the
requirements for CIV operability currently in the TS. the Limiting
Condition for Operation and the Surveillance Requirements would be
retained in the revised TS. Therefore, the proposed changes will not
affect the meaning, application, and function of the current TS
requirements for the CIVs.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: September 25, 1996
Description of amendment request: The amendments would change
Technical Specification 3/4.8.1, ``Electrical Power Systems,'' to
revise the Emergency Diesel Generator (EDG) voltage and frequency
limits as a result of updated EDG load calculations and to eliminate
ambiguity in the testing methodology for EDG start timing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Since no change is being made to the offsite power supplies, or
to any system or component that interfaces with the offsite power
supplies, there is no change in the probability of a Loss of Offsite
Power Accident.
The proposed changes provide the necessary conservatism for
voltage and frequency to ensure the EDGs are not run in an
overloaded condition and that driven equipment is not damaged during
steady state operation following a Loss of Offsite Power coincident
with a Loss of Coolant Accident. Since the narrower band of voltage
and frequency for the isochronous mode continues to ensure proper
steady state operation of the EDG and associated driven equipment,
there is no change in the consequences of an accident previously
evaluated.
Based on the above, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment does not result in any design or physical
configuration changes to the EDGs. Proposed changes made to the
testing parameters and testing methodology will not cause a new or
different accident since the EDGs are used for accident mitigation
and no new failure modes are being introduced. Therefore, the
proposed amendment will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed amendment provides further conservatism to the
voltage and frequency band currently specified in the TSs. The
proposed voltage and frequency changes ensure the EDG will not be
overloaded from an over-frequency condition and driven equipment
will not be damaged from an over-voltage condition.
The control system is set to control the EDG voltage within the
bands specified in the requested changes. The changes are consistent
with current calculations and within the capability of the controls.
Since the narrower band of voltage and frequency for the isochronous
mode is bounded by the existing TS, there is no change in the margin
of safety. The increased band for droop mode will ensure the EDG is
capable of operating in accordance with normal offsite power
parameters and does not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
[[Page 55040]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: October 1, 1996
Description of amendment request: The proposed amendments would
change Technical Specifications (TSs) 3/4.7.1.5, ``Main Steam Line
Isolation Valves (MSIVs),'' and 3/4.3.2, ``Engineered Safety Feature
Actuation System Instrumentation.'' These changes are needed to
accommodate entry into Modes 3 and 2 prior to performing MSIV closure
time testing in Mode 2. The proposed amendments would also allow for
the repair and testing of inoperable MSIVs in certain operating Modes,
and would change the low steam line pressure trip setpoint value for
safety injection to make it consistent with the previously approved
value for steam line isolation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The isolation capability of the MSIVs and the protective
functions of the low steam line pressure channels are necessary for
accident mitigation and do not impact the probability of an
accident. MSIV testing in the higher modes is necessary to obtain
conditions which enable testing of the MSIVs. These conditions are
consistent with the current accident analyses for main steam line
breaks and secondary system depressurization. Failure of a MSIV,
which could be encountered during testing, is accounted for in the
accident analyses.
Provisions for entering Mode 2 within six hours with an
inoperable MSIV allows operators to remove the plant from power
generation in a more controlled manner without challenging plant
safety systems and is consistent with other plant shutdown TS (i.e.,
TS 3.0.3). The additional six hours to Hot Shutdown, should MSIV
closure be infeasible, does not result in a significant increase in
the probability or consequence of an accident since this is a very
small incremental time addition. The values for the low steam line
pressure safety injection are higher and are bounded by the present
accident analysis. The elimination of the obsolete stroke time of
eight seconds is editorial in nature. As a result, the changes
proposed do not involve a significant increase in the probability or
consequence of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve any modifications to
existing plant equipment, do not alter the function of any plant
systems, do not introduce any new operating configurations or new
modes of plant operation, nor change the safety analyses. The
proposed changes will, therefore, not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
MSIV testing in Mode 2 is within the currently analyzed plant
operation as discussed in the Updated Final Safety Analysis Report
(UFSAR) Sections 10.3 and 15.4. These UFSAR sections address
performance of the TS surveillance test at or near 1000 psig Steam
Generator pressure to assure main steam isolation occurs within the
accident conditions, where Steam Generator pressure may be lower
during Mode 1 operation. The test methodology demonstrating MSIV
operability is consistent with the accident analysis.
Operation in Modes 2 and 3 with one or more isolation valve
inoperable and in the closed position does not impact the margin of
safety since the valves are already performing the safety function.
The protective functions that occur as a result of the low steam
line pressure initiating signal remain bounded by the values assumed
in the safety analyses. That is, the protective functions that occur
as a result of this initiating signal already assume a setpoint that
is conservative for the revised value. The change to the setpoint
eliminates conflicting information in the TS.
Therefore, the proposed changes does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of amendment request: September 20, 1996, as supplemented
September 30, 1996
Description of amendment request: The proposed amendment would
change Technical Specification 4.7.7.b.4 to indicate that the specified
flowrate for the Auxiliary Building Exhaust Air Filtration System
applies only to system testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The accident considered in this proposed change is the Loss of
Coolant Accident (LOCA) as described in Section 15.4 of the UFSAR
[Updated Final Safety Analysis Report]. The assumption is that:
``The Auxiliary Building Ventilation System will discharge the vapor
(from recirculation liquid leakage) to the atmosphere through
charcoal filters which have an efficiency of 90 percent.'' As such
the system acts to limit the total offsite and control room
radiation doses following a LOCA.
The Auxiliary Building Ventilation System [ABVS] is designed to
maintain the Auxiliary Building at a negative pressure with respect
to the atmosphere during normal and emergency operation. Filtration
of radio-iodines is accomplished by administratively aligning the
ECCS [emergency core cooling system] equipment areas exhaust flows
to the standby charcoal adsorber bed if required. The ABVS has no
direct impact on reactor operation or on any system connected to the
Reactor Coolant Pressure Boundary.
The emergency operation of the Auxiliary Building Ventilation
System is not affected by the proposed changes. The acceptance
criteria for system performance are not modified by the requested
change. The change clarifies the intent of SR [surveillance
requirement] 4.7.7.b.4 and the basis for the flowrates used for
system acceptance testing. It has been determined that operation of
the system at lower flow rates than those specified for surveillance
testing is conservative with respect to the radio-iodine removal
efficiency assumed for the charcoal adsorber. A higher removal
efficiency results in lower total exposures at the site boundary and
within the control room. Additionally, the system is capable of
maintaining the required negative pressure at the reduced flowrate.
Given the above, it is concluded that the proposed change does
not result in an increase in the probability or consequences
associated with previously analyzed accidents.
[[Page 55041]]
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment does not result in any design or
operational change to the ABVS, to the Nuclear Steam Supply System,
to the ECCS System, to the Containment Building, to the fuel or to
the electrical power supplies. Therefore, the proposed amendment
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Specification 3/4.7.7 and the associated bases were reviewed to
determine if the proposed changes result in a reduction in the
margin of safety. The change to SR 4.7.7.b.4 continues to assure
that the system is operated consistent with the assumptions of the
accident analysis. The proposed changes to Bases 3/4.7.7 clarify the
basis for flowrates associated with ABVS surveillance test
requirements. All changes result in ABVS operation that is just as
conservative as that assumed in existing analyses.
The proposed changes do not involve the addition or modification
of plant equipment, are consistent with the design basis of the ABVS
as described in the UFSAR, and appropriately limit operation to be
consistent with the assumptions of the accident analysis. As such
there is no 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: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 17, 1996
Brief description of amendments request: The proposed amendments
would modify the technical specifications to change (1) the reference
method for calculating dose conversion factors (DCFs) to be used in
dose calculations, and (2) the upper and lower limits for operating
pressurizer pressure to account for new instrument uncertainties and to
reduce the allowed operating band.
Date of individual notice in Federal Register: September 11, 1996
(61 FR 47963)
Expiration date of individual notice: October 11, 1996
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 28, 1996
Brief description of amendments request: The proposed amendments
would modify the technical specifications to increase the minimum
required amount of anhydrous trisodium phosphate (TSP) in the
containment baskets.
Date of individual notice in Federal Register: September 11, 1996
(61 FR 47962), as corrected September 26, 1996 (61 FR 50535).
Expiration date of individual notice: October 11, 1996
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of application for amendment: August 23, 1996
Brief description of amendment request: The proposed amendment
would revise Paragraph 2.B(2) of
Facility Operating License No. DPR-40 to allow source materials in
the form of depleted or natural uranium as reactor fuel and to revise
Technical Specification 4.3.2 to include depleted uranium in describing
the reactor core.
Date of individual notice in Federal Register: August 30, 1996 (61
FR 45995)
Expiration date of individual notice: September 30, 1996
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of application for amendment: September 19, 1996
Brief description of amendment request: The proposed amendments
would change Technical Specification requirements related to the low
temperature overpressure protection (LTOP) system. Specifically, the
reactor coolant system (RCS) temperature below which LTOP is required
to be enabled and one high pressure safety injection pump is required
to be rendered inoperable would be changed from 275 deg.F to 355
deg.F. Also, a specification would be added stating that only one
reactor coolant pump shall be operated when the RCS temperature is less
than or equal to 125 deg.F. Finally, editorial changes would be made
to rename the ``Overpressure Mitigating System'' as the ``Low
Temperature Overpressure Protection System.'' Date of individual notice
in Federal Register: October 1, 1996 (61 FR 51308) Expiration date of
individual notice: October 31, 1996
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth, Two Rivers, Wisconsin 54241
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: September 27, 1996
Brief description of amendment request: The proposed amendment
would change Technical Specification (TS) requirements related to the
low temperature overpressure protection (LTOP) system. Specifically,
the LTOP curve would be modified to define 10 CFR Part 50, Appendix G
pressure temperature limitations for LTOP evaluation through the end of
operating cycle (EOC) 33. In addition, the LTOP enabling temperature
and the temperature required for starting a reactor coolant pump would
be changed consistent with the design basis for the LTOP system.
Finally, the TS bases would be changed consistent with he changes
described above.
Date of individual notice in Federal Register: October 7, 1996 (61
FR 52472)
[[Page 55042]]
Expiration date of individual notice: November 6, 1996
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Notice Of Issuance Of Amendments ToFacility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: July 19, 1996
Brief description of amendment: The amendment revises the
containment spray nozzle surveillance interval in TS 3/4.6.2 from 5 to
10 years.
Date of issuance: October 3, 1996
Effective date: October 3, 1996
Amendment No.: 67
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44354) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 3, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas
Date of amendment request: April 11, 1996, as supplemented August
23, 1996
Brief description of amendments: The amendments revised the
Technical Specifications to permit implementation of 10 CFR Part 50,
Appendix J, Option B.
Date of issuance: October 3, 1996
Effective date: October 3, 1996
Amendment Nos.: 185 and 176
Facility Operating License Nos. DPR-51 and NPF-6: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20846) The additional information contained in the supplemental letter
dated August 23, 1996, 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 amendments is contained in a Safety
Evaluation dated October 3, 1996.No significant hazards consideration
comments received: No.
Public Document Room location: Tomlinson Library, Arkansas Tech
University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: April 29, 1996
Brief description of amendment: The amendment relocated cycle
specific operating parameters from the Technical Specifications to the
Core Operating Limits Report per Generic Letter 88-16. The parameters
being relocated by this amendment include the variable low reactor
coolant system pressure trip and the variable low reactor coolant
system pressure-temperature protective limits.
Date of issuance: October 3, 1996
Effective date: October 3, 1996
Amendment No.: 186
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28613) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 3, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 7, 1995, as supplemented by
letter dated April 11, 1996.
Brief description of amendment: The amendment modifies the Appendix
A Technical Specifications related to Safety Injection Tank level and
pressure setpoints.
Date of issuance: September 27, 1996
Effective date: September 27, 1996
Amendment No.: 121
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58401) The additional information contained in the supplemental
letter dated April 11, 1996, 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 September 27, 1996.No significant hazards consideration comments
received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: July 17, 1996
Brief description of amendments: The amendments consist of changes
to the Technical Specifications regarding containment leakage tests.
Date of issuance: October 4, 1996
Effective date: October 4, 1996
Amendment Nos.: 192 and 186Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44357)
[[Page 55043]]
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated October 4, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: May 21, 1996
Brief description of amendments: The amendments revise the
condensate storage tank level indication to ensure that the water level
is sufficient to provide 50,000 gallons of water for core spray makeup
to the reactor pressure vessel. On September 24, 1996, based on a
teleconference between the licensee and the NRC project manager, it was
mutually agreed to change the requested implementation schedule from 90
days to 30 days.
Date of issuance: October 2, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 202 and 143
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44358) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 2, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island
Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania
Date of application for amendment: January 16, 1995
Brief description of amendment: This amendment revised the
Technical Specification to incorporate an improvement from
administrative controls section of the revised standard TS for B&W
plants.
Date of issuance: October 8, 1996
Effective date: October 8, 1996
Amendment No.: 50Possession-Only License No. DPR-73: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65679). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 8, 1996No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: July 5, 1996
Brief description of amendment: The amendment will support the
implementation of noble metal chemical addition at the Duane Arnold
Energy Center as a method to enhance the effectiveness of hydrogen
water chemistry in mitigating intergranular stress corrosion cracking
in reactor vessel internal components. Specifically, the amendment will
permit an increase of the reactor water conductivity limit in Technical
Specification (TS) Table 3.6.B.2-1 and several other changes in TS
sections 4.6.B.2.c, 4.6.B.2.d, and the associated Bases.
Date of issuance: October 3, 1996
Effective date: October 3, 1996
Amendment No.: 218
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40020) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 3, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: December 22, 1995, as
supplemented September 20, 1996
Brief description of amendment: The amendment revises the Duane
Arnold Energy Center (DAEC) Technical Specifications (TS) Sections
3.7.A and 4.7.A, ``Primary Containment,'' by deleting information also
contained in 10 CFR Part 50, Appendix J, Option A and incorporating
references to the Primary Containment Leakage Rate Testing Program.
These changes allow the use of the performance based option of
containment leak testing. The amendment also adds Operability and
Surveillance Requirements (SRs) for the drywell air lock. Minor
administrative changes were also made. These changes are consistent
with comparable specifications in the Improved Standard Technical
Specifications (ITS), NUREG-1433. In addition, the staff executed
administrative changes and corrections to the TS Bases, as submitted in
two letters dated February 13, 1995. Sections changed or corrected are
Section 1.2, Bases; Section 2.2, Bases Reactor Coolant System
Integrity; Section 3.7.H/4.7.H, Bases Containment Atmosphere Dilution;
and Section 3.7.I/4.7.I, Bases Oxygen Concentration.
Date of issuance: October 4, 1996
Effective date: October 4, 1996
Amendment No.: 219
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3499) The September 20, 1996, submittal was clarifying in nature and
did not affect the no significant hazards determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 4, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: June 28, 1996 and as
supplemented on September 17, 1996
Brief description of amendment: The amendment will allow removal of
the Inclined Fuel Transfer System (IFTS) primary containment blind
flange while primary containment is required to be operable. This will
provide flexibility to operate the IFTS for the purpose of testing and
exercising the system during such conditions. Primary containment
integrity will be provided by an alternate means while the blind flange
is removed. The change will be incorporated via a provisional note into
Technical Specification (TS) Surveillance Requirement 3.6.1.3.3,
associated with TS 3.6.1.3, ``Primary Containment Isolation Valves
(PCIVs).''
Date of issuance: October 3, 1996
Effective date: October 3, 1996
Amendment No.: 107
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
[[Page 55044]]
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40021) The information provided in the licensee's letter of September
17, 1996 provided clarifying information and did not involve
significant changes to the original Federal Register notice.The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 3, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: June 21, 1996, and as
supplemented by letter dated August 15, 1996
Brief description of amendment: The amendment modifies Section 5.7,
``High Radiation Areas,'' of the ``Administrative Controls'' section of
the Clinton Power Station technical specifications (TS). The changes
include: (1) allowing utilization of a Radiation Work Permit (RWP) ``or
equivalent'' to control entry into a high radiation area; (2)
clarifying the example given in the TS of individuals who are qualified
in radiation protection procedures; (3) clarifying the requirements for
when specified access controls and barriers for high radiation areas
within large areas like the containment may be established; (4)
clarifying that it is acceptable for an RWP to specify a maximum dose,
i.e., a specified setpoint on an alarming dosimeter in lieu of a stay
time for entry into a high radiation area (where an individual could
receive a deep dose equivalent of 3000 mrem in one hour); (5)
eliminating the upper dose limit for specifying the applicability of
the requirements of Specification 5.7.1; (6) providing additional
flexibility regarding the control of keys to locked doors for
preventing unauthorized entry into high radiation areas; (7) providing
alternate means of informing individuals of dose rates in immediate
work areas; (8) reorganizing TS Sections 5.7.1, 5.7.2, and 5.7.3 into
four sections (5.7.1, 5.7.2, 5.7.3 and 5.7.4); and (9) making minor
edits to enhance readability.
Date of issuance: October 3, 1996
Effective date: October 3, 1996
Amendment No.: 108
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40021) The August 21, 1996, submittal consisted of supporting technical
information which did not change the staff's initial proposed no
significant hazards consideration determination or expand the scope of
the original notice. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated October 3, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: May 2, 1996, as supplemented by
letter dated August 30, 1996
Brief description of amendment: The amendment removes Technical
Specification Figure 5.1, which was used in maintaining Keff
values, and substitutes in its place a defined requirement for maximum
Kinfinity for any fuel placed in the Millstone Unit 1 spent fuel
pool.
Date of Issuance: October 4, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 97
Facility Operating License No. DPR-21: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 17, 1996 (61 FR
37301) The August 30, 1996, letter provided additional, clarifying
information that did not change the scope of the May 2, 1996,
application and the initial proposed no significant hazards
consideration determination.The Commission's related evaluation of this
amendment is contained in a Safety Evaluation dated October 4, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, Connecticut
Northern States Power Company, Docket No. 50-282, Prairie Island
Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota
Date of application for amendment: July 15, 1996, and supplemented
August 22, 1996
Brief description of amendment: The amendment allows the use of the
moveable in-core detector system for measurement of the core peaking
factors with less than 75 percent and greater than or equal to 50
percent of the detector thimbles available. The amendment is a one-time
only change for Prairie Island, Unit 1, to reduce the number of
required in-core detectors necessary for continued operation for the
remainder of Operating Cycle 18.
Date of issuance: October 10, 1996
Effective date: October 10, 1996, and shall remain effective for
the remainder of Cycle 18 only
Amendment No.: 124
Facility Operating License No. DPR-42. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40024) By letter dated August 22, 1996, NSP forwarded a copy of the
results of its most recent low power physics tests to the NRC for use
as a reference and provided additional clarifying information. This
information was within the scope of the original application and did
not change the staff's initial proposed no significant hazards
considerations determination. Therefore, renoticing was not
warranted.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 10, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: May 17, 1996
Brief description of amendment: The amendment revises Technical
Specifications (TS) 2.18, 3.14, 3.3, and 5.10 to relocate the
operability requirements for shock suppressors (snubbers) from the TS
to the Updated Safety Analysis Report (USAR) and incorporate snubber
examination and testing requirements in TS 3.3.
Date of issuance: September 27, 1996
Effective date: September 27, 1996
Amendment No.: 176
[[Page 55045]]
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44360) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 27, 1996.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: August 23, 1996
Brief description of amendment: The amendment modifies paragraph
2.B.(2) of
Facility Operating License No. DPR-40 allowing the use of source
material, in the form of depleted or natural uranium, as reactor fuel.
Date of issuance: October 2, 1996
Effective date: October 2, 1996
Amendment No.: 177
Facility Operating License No. DPR-40: Amendment revised the
Operating License.
Date of initial notice in Federal Register: August 30, 1996 (61 FR
45995) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: January 25, 1996
Brief description of amendment: The amendment would extend the
instrumentation surveillance test intervals to support 24-month
operating cycles. These proposed changes would eliminate the mid-cycle
outages to perform the Technical Specification surveillance
requirements.
Date of issuance: October 2, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 233
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25709) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: March 27, 1996, as supplemented
April 24, 1996, and August 15, 1996
Brief description of amendment: The proposed amendment changes
would permit implementation of 10 CFR Part 50, Appendix J, Option B
with an exception to the guidelines of Regulatory Guide 1.163 for Type
C testing of primary containment isolation valves in the reverse (non-
accident) direction.
Date of issuance: October 4, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 234
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20855) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 4, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: August 9, 1996, as supplemented
September 17, 1996
Brief description of amendment: The amendment revises the Technical
Specifications to revise the safety limit minimum critical power ratio
for cycle 19 operation from its current value of 1.07 (for the fuel
currently in the reactor for cycle 18) for two recirculation loop
operation to 1.10, and from 1.08 to 1.12 for single recirculation loop
operation.
Date of issuance: October 4, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 150
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44364) The September 17, 1996, letter provided clarifying information
that did not change the scope of the August 9, 1996, application and
initial proposed no significant hazards consideration determination.The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 4, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: April 16, 1996, as supplemented
July 25, 1996
Brief description of amendment: The amendment permits
implementation of 10 CFR Part 50, Appendix J, Option B, ``Performance-
Based Requirements.''
Date of issuance: October 2, 1996
Effective date: October 2, 1996
Amendment No.: 135
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34898) The July 25, 1996, supplement provides clarifying information
and did not change the scope of the initial notice. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated October 2, 1996.No significant hazards consideration comments
received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Southern California Edison Company, et al, Docket No. 50-206, San
Onofre Nuclear Generating Station, Unit No. 1, San Diego County,
California
Date of application for amendment: March 13, 1996
Brief description of amendment: The change revises the San Onofre
Unit 1 License Condition 2.D. This change eliminates a reporting
requirement that is redundant to reporting requirements in 10 CFR 50.72
and 50.73. Additionally, the amendment makes administrative and
editorial changes to the Permanently Defueled Technical Specifications.
Date of issuance: October 3, 1996
[[Page 55046]]
Effective date: October 3, 1996
Amendment No.: 158
Facility Operating License No. DPR-13: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40028) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 3, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Science Library, University of
California, Irvine, California 92713
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: December 6, 1995, as
supplemented by letters dated August 30, 1996, and September 20, 1996
Brief description of amendments: These amendments revise Technical
Specifications (TS) Section 4.3 ``Fuel Storage'' to allow fuel
assemblies having a maximum U-235 enrichment of 4.8 weight percent (w/
o) to be stored in both the spent fuel racks and the new fuel racks.
Additionally, TS Section 3.7.18 ``Spent Fuel Assembly Storage,''
Figures 3.7.18-1 ``Unit 1 Fuel Minimum Burnup vs. Initial Enrichment
for Region II Racks,'' and 3.7.18-2 ``Units 2 and 3 Fuel Minimum Burnup
vs. Initial Enrichment for Region II Racks,'' are being revised and
relabeled.
Date of issuance: October 3, 1996
Effective date: October 3, 1996, to be implemented within 30 days
as of the date of issuance.
Amendment Nos.: Unit 2 - 131; Unit 3 - 120
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15997) The August 30, 1996, and September 20, 1996, letters provided
additional clarifying information and did not change the initial no
significant hazards consideration determination.The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated October 3, 1996.No significant hazards consideration
comments received: No.
Temporary Local Public Document Room location: Science Library,
University of California, P. O. Box 19557, Irvine, California 92713
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: July 31, 1996 (TXX-96433)
Brief description of amendments: The amendments revised core safety
limit curves (Technical Specification (TS) Figure 2.1-1a) and new N-16
setpoint values and parameters (TS Table 2.1-1) for Unit 1, and
reference to topical report RXE-95-001-P as an approved methodology for
small break loss of coolant accident analysis for Units 1 and 2.
Date of issuance: September 30, 1996
Effective date: September 30, 1996, to be implemented within 30
days
Amendment Nos.: 52 and 38
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44362) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 30, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: April 12, 1996, as supplemented
by letters dated August 2, 1996, August 19, 1996, and September 5,
1996.
Brief description of amendment: The amendment revises the Technical
Specifications to address the installation of laser welded tube sleeves
in the Callaway Plant steam generators.
Date of issuance: October 1, 1996
Effective date: October 1, 1996, and will be implemented within 30
days of the date of issuance.
Amendment No.: 116
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20857) The August 2, 1996, August 19, 1996, and September 5, 1996,
supplemental letters provided clarifying information and did not change
the original no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluationdated October 1, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: April 17, 1996, as supplemented
by letters dated July 15, 1996, July 31, 1996, and August 28, 1996.
Brief description of amendment: The amendment would change
Technical Specification (TS) 3/4.3 to support a future modification to
replace existing digital portions of the main steam and feedwater
isolation system (MSFIS) with digital processor equipment and would
authorize revision of the FSAR to include a description of the MSFIS
modification.
Date of issuance: October 1, 1996
Effective date: October 1, 1996, to be implemented prior to startup
from the Callaway Plant Refuel 8.
Amendment No.: 117
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications and the Final Safety Analysis Report.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28619) The July 15, 1996, July 31, 1996 and August 28, 1996
supplemental letters provided additional clarifying information and did
not change the staff's original no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 1, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: April 4, 1996
Brief description of amendment: The amendment revises the Technical
Specifications regarding the surveillance requirement for control rod
over-travel by moving the specific testing methodology to licensee
administratively controlled documents. Specifically, the amendment
removes the requirement in Specification 4.3.B.1(b) to verify prior to
coupling that the over-travel indicating light is working properly by
withdrawing an uncoupled control rod drive to the over-travel position.
[[Page 55047]]
Date of issuance: September 30, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 149
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20860) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 1996.No
significant hazards consideration comments received: No
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: August 9, 1996
Brief description of amendment: The amendment changes the
operations manager qualification requirements to allow either of two
alternatives (having held a senior reactor operator's license or having
been certified for equivalent senior reactor operator knowledge) to the
requirement for the operations manager to hold a senior reactor
operator's license.
Date of issuance: October 1, 1996
Effective date: October 1, 1996
Amendment No.: 148
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44350) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 1, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: July 3, 1996, as supplemented on
July 23, August 28, and September 16, 1996
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant Technical Specification 4.2.b, ``Steam Generator
Tubes,'' and its associated basis, by revising the acceptance criteria
for indications of tube degradation occurring in the tubesheet crevice
region.
Date of issuance: October 2, 1996
Effective date: October 2, 1996
Amendment No.: 129
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40031) The July 23, August 28, and September 16, 1996, submittals
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination.The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 2, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
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 application for amendments: May 29, 1996, as supplemented
August 20, 1996
Brief description of amendments: These amendments revise Technical
Specification (TS) Section 15.4.4, ``Containment Tests,'' to
incorporate the provisions of 10 CFR Part 50, Appendix J, ``Primary
Reactor Containment Leakage Testing for Water-Cooled Power Reactors,''
Option B. Revisions have also been made to TS Sections 15.1,
``Definitions,'' 15.3.6, ``Containment System,'' and 15.6,
``Administrative Controls,'' to support the proposed changes to Section
15.4.4.
Date of issuance: October 9, 1996
Effective date: October 9, 1996, to be implemented within 45 days.
Amendment Nos.: Unit 1 - 169 and Unit 2 - 173
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34901) The supplemental information did not affect the staff's initial
no significant hazards consideration determination.The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated October 9, 1996.No significant hazards consideration
comments received: No
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
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
[[Page 55048]]
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, 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 November 22, 1996, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. 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 20555-001, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-001, 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).
[[Page 55049]]
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 21, 1996
Brief description of amendments: The amendments approve changes to
the Updated Final Analysis Report (UFSAR), and require that the changes
be submitted with the next update of the UFSAR pursuant to 10 CFR
50.71(e). The associated Safety Evaluation delineates the staff's
review and findings, including finding that the as-built condition of
the subject power system protective devices is acceptable as-is.
Date of issuance: September 28, 1996
Effective date: September 28, 1996
Amendment Nos.: 153 and 145
Facility Operating License Nos. NPF-35 and NPF-52: The amendments
revised the Updated Final Safety Analysis Report. Public comments
requested as to proposed no significant hazards consideration: Yes. The
NRC staff published a public notice of the proposed amendments, issued
a proposed finding of no significant hazards consideration, and
requested that any comments on the proposed no significant hazards
consideration be provided to the staff no later than 5:00 p.m.,
September 28, 1996. The notice was published in ``The Herald'' of Rock
Hill, South Carolina, from September 25 through 27, 1996. No comments
have been received.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, consultation with the State of South Carolina,
and final determination of no significant hazards consideration are
contained in a Safety Evaluation dated September 28, 1996.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit
No. 2, York County, Pennsylvania
Date of application for amendment: March 25, 1996 as supplemented
by letters dated August 23, 1996 and September 27, 1996.
Brief description of amendment: The amendment revises Peach Bottom
Technical Specification 2.1.1.2 safety limit minimum critical power
ratios to be consistent with the use of GE-13 fuel in the Unit 2 core
for operating cycle 12.
Date of issuance: September 27, 1996
Effective date: As of date of issuance
Amendment No.: 217
Facility Operating License No. DPR-44: Amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: Yes (61 FR 45997). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided an opportunity to request a
hearing by September 30, 1996, but indicated that if the Commission
makes a final no significant hazards consideration determination any
such hearing would take place after issuance of the amendment.The
Commission's related evaluation of the amendment, finding of exigent
circumstances, and final no significant hazards consideration
determination are contained in a Safety Evaluation dated September 27,
1996.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. Vice
President and General Counsel, PECO Energy Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Dated at Rockville, Maryland, this 16th day of October 1996.
For the Nuclear Regulatory Commission
John A. Zwolinski,
Acting Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[FR Doc. 96-27025 Filed 10-22-96; 8:45 am]
BILLING CODE 7590-O1-F