[Federal Register Volume 60, Number 177 (Wednesday, September 13, 1995)]
[Notices]
[Pages 47613-47630]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-22616]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating
LicensesInvolving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 18, 1995, through August 30, 1995.
The last biweekly notice was published on Wednesday, August 30, 1995
(60 FR 45172).
[[Page 47614]]
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By October 13, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public
[[Page 47615]]
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: July 17, 1995
Description of amendment request: The requested change to Technical
Specification (TS) section 3.8 would specify that the spent fuel
building refueling filter fan and at least one containment purge fan
shall be shown to operate within plus or minus 10 percent of the design
flow.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:The proposed change to TS is to
revise Section 3.8.2.c. This TS section currently states ``All filter
system fans shall be shown to operate within [plus or minus] 10% of the
design flow.'' The proposed requirements are as follows:
c.1 The Spent Fuel Building refueling filter fan shall be shown
to operate within [plus or minus] 10% of the design flow.
c.2 At least one Containment purge filter fan shall be shown to
operate within [plus or minus] 10% of the design flow and must be
operable during core alterations or movement of irradiated fuel
assemblies, or at least one automatic containment isolation valve in
each line penetrating the containment which provides a direct path
from the containment atmosphere to the outside atmosphere shall be
securely closed.
This proposed change does not involve a significant hazards
consideration for the following reasons.
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change clarifies the operating requirements
for the Containment purge and Spent Fuel Building refueling filter
systems. This proposed change to the TS specifically delineates the
fan filter systems required for refueling operations and does not
change the physical operation of the filter systems. The affected
systems are not involved in the initiation of any accident. The
system response to previously analyzed accidents, including system
flows and filter efficiencies will not be altered by the proposed
change. These changes are enhancements to clarify existing TS
requirements that will not increase the probability or consequences
of a previously analyzed accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed change merely clarifies the specific filter
systems that are necessary to mitigate a fuel handling accident
during core alterations or the movement of irradiated fuel
assemblies and is consistent with the accident analysis in Section
15.7.4 of the Updated Final Safety Analysis Report (UFSAR). This
proposed change does not involve the addition or modification of
plant equipment, nor does it alter the design or operation of plant
systems. Therefore, operation of the facility in accordance with the
proposed TS change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change clarifies which filter systems that must
be capable of mitigating a design basis fuel handling accident
during core alterations or the movement of irradiated fuel
assemblies and is consistent with the accident analysis in Section
15.7.4 of the UFSAR. The proposed change will not result in an
increase in the Control Room or offsite radiation doses. The
performance of the filtration systems, including adsorption
efficiencies, will not change. Therefore, the proposed change 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: Hartsville Memorial Library,
147 West College Avenue, Hartsville, SC 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, NC 27602
NRC Project Director: David B. Matthews
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: June 30, 1995
Description of amendment request: The proposed amendments would
modify the emergency diesel generator testing requirements in the
Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of occurrence of any accident
previously evaluated.
The proposed changes to the Technical Specifications will change
the scope of EDG [Emergency Diesel Generator] testing that is
performed on a refueling cycle frequency. The proposed changes will
eliminate the requirement to perform sequenced loading of the EDG as
part of the hot restart test, and will allow the hot restart test to
be initiated from any EDG start signal. The revised requirements
will eliminate testing that is redundant, provides no additional
meaningful information, significantly constrains scheduling of
refueling outage maintenance and testing, and impacts the
availability of systems and components important to safety. The
proposed testing requirements satisfy the underlying purpose of the
EDG hot restart test. The testing in accordance with the proposed
requirements will verify the ability of each EDG to complete the
start up sequence from an equilibrium temperature immediately
following operation at full load for a period of time long enough to
stabilize operating temperature.
A two hour period for operation at full load has been chosen to
ensure that full load operating temperature has stabilized prior to
shutdown preceding the hot restart test. Momentary transients
outside the full load operating band of 3600 to 4000 kW will not
invalidate the two hour run since momentary transient will not
significantly affect operating temperature. Brief operation
subsequent to a momentary transient will normalize operating
temperature. Since the proposed changes impact only surveillance
requirements used to periodically verify the operability of a
required safety system, and since the proposed changes provide an
[[Page 47616]]
equivalent level of testing and eliminate redundant testing, the
proposed changes will not impact the operability or availability of
a required system.
Operation in accordance with the revised requirements will not
increase the likelihood that a transient initiating event will occur
since transients are initiated by equipment malfunction and/or
catastrophic system failure. The revised requirements affect testing
that is performed on a Refueling Cycle frequency. Testing in
accordance with the proposed requirements will not increase the
probability of failure of the EDGs since the testing will provide an
equivalent level of testing to verify the operability of the EDGs.
In addition, failure of an EDG to start or failure of an EDG while
operating is not assumed to be an initiating event of an accident
considered in the Updated Final Safety Analysis Report (UFSAR).
Based on the above, operation in accordance with the proposed
requirements will not significantly increase the probability of
occurrence of any accident previously evaluated.
The proposed requirements will meet the underlying purposed of
the existing testing requirements. The proposed testing will ensure
the ability of the EDG to start from a hot condition in the unlikely
event of an accident. The proposed changes will eliminate testing
requirements that are redundant and unnecessarily challenge the
reliability of the EDGs by requiring unnecessary wear and cycling of
the diesel engine and auxiliary systems. Since the proposed changes
will not adversely affect the operability or availability of the
EDGs, the ability of the EDGs to operate and power equipment
important to safety will not be impacted and the ability to mitigate
the consequences of accidents previously evaluated will not be
affected. Based on the preceding discussion, the consequences of
accidents previously evaluated will not significantly increase.
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 to the Technical Specifications do not
involve the addition of any new or different types of safety related
equipment, nor do they involve the operation of equipment required
for safe operation of the facility in a manner different from those
addressed in the UFSAR. No safety related equipment or function will
be altered as a result of the proposed changes. Also, the procedures
that govern normal operation and recovery from an accident are not
affected by the proposed changes.
The proposed changes will eliminate testing requirements that
are redundant and provide no additional meaningful information.
Testing in accordance with the revised requirements will provide an
equivalent level of confidence in the reliability of the EDG systems
to complete the start up sequence from a hot condition. The proposed
testing requirements satisfy the purpose Regulatory Guide 1.108 in
that the testing requirements will ensure EDG operability and
reliability. In addition, the proposed changes are consistent with
the changes recommended by the NRC in Generic Letter 93-05. Since no
new failure modes or mechanisms are introduced by the proposed
changes, the possibility of a new or different kind of accident is
not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Plant safety margins are established through LCOs, limiting
safety system settings, and safety limits specified in the Technical
Specifications. There will be no changes to either the physical
design of the plant or to any of these settings or limits as a
result of the proposed changes. The proposed changes will eliminate
testing requirements that are redundant and provide no additional
information. Testing in accordance with the revised requirements
will verify the ability of the EDGs to complete the start up
sequence from a hot condition as is intended by the recommended
testing in Regulatory Guide 1.108. In addition, the proposed changes
are consistent with the changes recommended by the NRC in Generic
Letter 93-05. Since the proposed changes will not impact the
availability or operability of the EDGs to perform their intended
function and since no LCOs, safety limits, or safety system settings
are affected by the proposed changes, there is no 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, IL 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, IL 60603
NRC Project Director: Robert A. Capra
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 26, 1995.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications to allow rod
misalignment of +/- 18 steps at or below 90% of rated thermal power. In
addition, a change is proposed to the Limiting Condition for Operation
range of rod travel from 228 to ``All Rods Out.'' The introduction of
``All Rods Out'' is consistent with Amendment 167/161 which approved
the removal of Technical Specification 3.1.3.6, ``Rod Insertion Limit''
from the Technical Specifications and placement into the Core Operating
Limits Report (COLR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of any accident previously evaluated.
The proposed limits on rod misalignment do not increase the
probability of an accident. The Technical Specifications' allowed
increase in peaking factor limits as power is reduced accommodates
an increase in rod misalignment of [plus or minus] 18 steps below
90% of RTP [rated thermal power]. The initial conditions remain
unchanged from that assumed in the Updated Final Safety Analysis
Report (UFSAR). Therefore, this proposed change poses no significant
increase in the probability or consequences of any accident
previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
No new accident scenarios, failure mechanisms or limiting single
failure are introduced as a result of implementing the proposed rod
misalignment criteria. The institution of the proposed rod
misalignment criteria will have no adverse effect, nor does it
challenge, the performance of any other safety related system.
Therefore, the proposed amendment does not in any way create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety. The margin of safety, as defined in the BASES for the
Technical Specifications, is not significantly affected by the
changes to the rod misalignment limit. The Technical Specifications'
allowed increase in peaking factor limits as power is reduced
accommodates an increase in rod misalignment of [plus or minus] 18
steps below 90% of RTP. The initial conditions remain unchanged from
that assumed in the UFSAR. Since the peaking factor limits are not
modified, the proposed change does not constitute 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: Florida International
University, University Park, Miami, FL 33199
[[Page 47617]]
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 26, 1995.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications to delete the
requirement to adjust the Nuclear Instrumentation System (NIS) downward
when operating at less than 70% of rated thermal power (RTP).
At reduced power levels (i.e., less than 70% of RTP), calorimetric
power measurement uncertainties are most influenced by the feedwater
flow measurements, which have the potential for large flow
uncertainties under low flow conditions. These calorimetric
uncertainties create the potential for a non-conservative gain
adjustment of the NIS when the NIS is adjusted downward to match
calorimetric power at reduced power levels, and may result in a non-
conservative NIS power level indication when operating at higher power
levels. Inappropriate gain adjustments could cause the NIS Power Range
High Neutron Flux trip to occur at power levels beyond that assumed in
the plant safety analyses. The proposed changes would correct this
situation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change does not involve any physical changes to the
NIS. Implementation of the proposed change does not affect the
probability of failure of the NIS and does not alter the method in
which protection is afforded by the NIS for the reactor and primary
system. Therefore, the proposed change does not result in an
increase in the severity or consequences of any accident previously
evaluated.
The proposed change in Technical Specifications to remove the
requirement which could result in non-conservative gain adjustments
of the NIS at reduced power levels (below 70% of RTP), will have no
significant effect on the probability or consequences of licensing
basis events; and the probability or consequences of an accident
previously evaluated for Turkey Point has not been significantly
increased. Therefore, operation of the facility in accordance with
the proposed amendments would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change does not result in a change in the method in
which the NIS provides plant protection. No change is being made
which alters the function of the NIS. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident nor involve a reduction in a margin of safety as defined in
the Safety Analysis Report.
The change in Technical Specifications associated with the
removal of the requirement which could result in non-conservative
gain adjustments of the NIS at reduced power levels (below 70% of
RTP) will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
This change in Technical Specifications only affects the removal
of the requirement which has the potential for non-conservative gain
adjustments of the NIS at reduced power levels (below 70% of RTP);
these changes do not alter the manner in which protection is
afforded for the reactor and primary system. In addition, the
fundamental process for implementation of the calorimetric power/NIS
comparison remains the same.
The changes in Technical Specifications associated with the
removal of the requirement, which could lead to non-conservative
gain adjustments of the NIS at reduced power levels (below 70% of
RTP), will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Florida International
University, University Park, Miami, FL 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 26, 1995.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) to incorporate
certain changes which are consistent with guidance provided by NUREG-
1366 and NRC Generic Letter (GL) 93-05, ``Line-Item Technical
Specification Improvements to Reduce Surveillance Requirements for
Testing During Power Operation.'' The following proposed changes are
requested:
(1) TS SR 4.1.3.1.2: Change the frequency interval for control rod
movement test from monthly to quarterly.
(2) TS SR 4.6.5.1: Change the hydrogen monitor calibration from
quarterly to each refueling interval, and the analog channel
operational test from monthly to quarterly.
(3) TS SR Table 4.3-3: Change the analog channel functional test
from monthly to quarterly for radiation monitors. Correct spelling of
'Radioactivity' in Item 1.a.
(4) TS SR 4.4.6.2.2: Increase the time allowed in COLD SHUTDOWN
before leak testing the Reactor Coolant System (RCS) isolation valves
is required, from 72 hours to 7 days.
(5) TS SR 4.10.1.2: Changes the requirement for a rod drop test
prior to reducing SHUTDOWN MARGIN from ``within 24 hours'' to ``within
7 days''.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed amendments conform to the guidance given in
Enclosure 1 of the NRC Generic Letter 93-05. The overall functional
capabilities of the rod control system, RCS pressure isolation
valves, the hydrogen monitoring system, and the radiation monitoring
systems will not be modified by the proposed change. These
amendments will not involve a significant increase in the
probability or consequences of an accident previously evaluated for
the following reasons:
(1) Increasing the interval of control rod movement testing will
reduce the possibility of testing-related reactor trips and dropped
rods, and result in fewer challenges to safety systems and plant
transients.
(2) Increasing the interval of hydrogen monitor calibration and
operational tests will result in a reduction in equipment
degradation and reduce a burdensome task on personnel resources.
[[Page 47618]]
(3) Increasing the interval of radiation monitor functional
tests will result in less equipment degradation as well as reducing
the potential for testing-related isolations of the control room,
fuel handling building, auxiliary buildings, and various process
lines.
(4) Increasing the time allowed in COLD SHUTDOWN prior to leak
testing RCS isolation valves will permit plant personnel to focus on
short notice outage recovery and minimize personnel radiation
exposure. Since the methods and the acceptance criteria used for the
leak test are not altered, increasing the time from 72 hours to 7
days will not significantly alter the associated risk.
(5) Increasing the time required to perform rod tests prior to
reducing the SHUTDOWN MARGIN will result in only one rod drop test
vice two following a refueling outage, which will in turn reduce
plant transients and personnel resource requirements.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The use of the proposed changes to the TS can not create the
possibility of a new or different kind of accident from any accident
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
the surveillance interval changes and clarifications, since the
proposed changes do not involve the addition or modification of
equipment nor do they alter the design or operation of affected
plant systems.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems are unchanged by the proposed amendments. The proposed
changes to the TS which establish new or clarify old surveillance
intervals consistent with the NRC Generic Letter 93-05 line-item
improvement guidance do not significantly reduce any of the margins
of safety even though the number of surveillances is decreased.
These requested amendments are justified by the following reasoning
from NUREG-1366:
(1) The surveillances could lead to plant transients which would
challenge safety systems unnecessarily as in the cases of control
rod movement tests and post-refueling rod drop tests.
(2) The surveillances result in the unnecessary wear to
equipment as in the cases of the hydrogen and radiation monitor
surveillances.
(3) The surveillance result in radiation exposure to plant
personnel which is not justified by the safety significance of the
surveillances as in the case of the time requirement for leak-
testing RCS isolation valves when in COLD SHUTDOWN.
(4) The surveillances place an unnecessary burden on plant
personnel because the time required is not justified by the safety
significance of the surveillance, i.e. hydrogen monitor and post-
refueling rod drop tests.
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: Florida International
University, University Park, Miami, FL 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 26, 1995.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specification Administrative
Controls Section 6.9.1.7 to reflect the use of the Westinghouse NOTRUMP
model in the Small Break Loss-of-Coolant Accident (SBLOCA) analysis
used in determining the K(z) curve contained in the Core Operating
Limits Report (COLR). The following references would be added to
Section 6.9.1.7 (COLR) of the Administrative Controls section of Turkey
Point Units 3 and 4 TS: WCAP-10054-P-A, (proprietary) and
WCAP-10081-NP-A, (non-proprietary), ``Westinghouse Small Break ECCS
Evaluation Model Using the NOTRUMP Code'', October, 1985.'' WCAP-10054-
P-A Addendum 2, (proprietary), ``Addendum to the Westinghouse Small
Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection
into the Broken Loop and COSI Condensation Model'', August, 1994.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The modification to the current Section 6.9.1.7 of the
Administrative Controls section of the Turkey Point Technical
Specifications to include the references to WCAP-10054-P-A, ``Small
Break ECCS Evaluation Model Using the NOTRUMP Code'', and WCAP-
10054-P-A Addendum 2 for the COSI model, does not involve an
increase in the probability or consequences of an accident
previously evaluated. This modification to the Technical
Specification does not change the probability of occurrence
previously evaluated.
This change does not affect the integrity of the fission product
barriers utilized for mitigation of radiological dose consequences
as a result of an accident. The addition of the new methodology used
for Turkey Point uprating analysis does not change, degrade, or
prevent the response of safety related mitigation systems to
accident scenarios, as described in the Updated Final Safety
Analysis Report (UFSAR) Chapter 14. Therefore, the licensee
concluded that the probability or consequences of an accident
previously evaluated are not increased.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The modification to the current Section 6.9.1.7 of the
Administrative Controls section of the Turkey Point Technical
Specifications to include the references to WCAP-10054-P-A, ``Small
Break ECCS Evaluation Model Using the NOTRUMP Code'', and WCAP-
10054-P-A Addendum 2 for the COSI model, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new operating configuration is being
imposed by the addition of the references to the Technical
Specification. Therefore, no new failure modes or limiting single
failures have been identified. The licensee concludes that no new or
different kind of accidents from those previously evaluated have
been created as a result of this revision.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
The modification to the current Section 6.9.1.7 of the
Administrative Controls section of the Turkey Point Technical
Specifications to include the references for the Small Break ECCS
Evaluation Model Using the NOTRUMP Code will not involve a reduction
in the margin of safety. The SBLOCA analysis results show that the
limits of 10 CFR 50.46 are maintained as follows. The new calculated
value of worst-case PCT will be 1688 deg.F, which is less than the
limit of 2200 deg.F. There is significant margin in the current
SBLOCA analysis such that the total cladding oxidation limit of 17
percent will not be challenged. Further, the calculated total amount
of hydrogen generated has been determined to remain less than 1
percent. The SBLOCA hydraulic forces are not affected by the K(z)
curve and the licensee concludes that the core will remain amenable
to cooling. Additionally, post-LOCA long term core cooling and hot
leg switchover evaluations are not impacted by the K(z) curve.
Therefore, there is 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request
[[Page 47619]]
involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, FL 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 26, 1995.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications to achieve
consistency throughout these documents by (a) removing outdated
material, (b) incorporating administrative clarifications and
corrections, and (c) correcting typographical errors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed amendments are purely administrative in nature.
These amendments will not involve a significant increase in the
probability or consequences of an accident previously evaluated
because they do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. Therefore, the proposed changes do not affect the
probability or consequences of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The use of the modified specifications can not create the
possibility of a new or different kind of accident from any
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
the administrative changes and clarifications, since the proposed
changes do not involve the addition or modification of equipment nor
do they alter the design or operation of affected plant systems,
structures, or components.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. The modified specifications which correct administrative
errors and clarify existing Technical Specification requirements do
not significantly reduce any of the margins 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: Florida International
University, University Park, Miami, FL 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 17, 1995
Description of amendment request: The proposed amendment would
allow the containment to be opened after about 11 days following
shutdown during refueling and would redefine the operability
requirements for selected engineered safety feature systems such that
these systems are only required to be operable during the calculated
decay period. The proposed changes will not remove requirements for
systems to mitigate potential vessel draindown events, will not remove
requirements for systems required for decay heat removal, and will
continue to require high water level over the vessel during fuel
movement. Programs are in place to close the containment, if needed, to
address shutdown risk concerns.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed limits on recently irradiated fuel is used to
establish operational conditions where specific activities represent
situations where significant radioactive releases can be postulated.
These operational conditions are consistent with the design basis
analysis. Because the equipment affected by the revised operational
conditions is not considered an initiator to any previously analyzed
accident, inoperability of the equipment cannot increase the
probability of any previously evaluated accident.
The proposed applicability in conjunction with existing
administrative controls on light loads, bounds the conditions of the
current design basis fuel handling accident analysis. The analysis
also concludes the limiting offsite radiological consequences are
well within the acceptance criteria of NUREG 0800, Section 15.7.4
and GDC 19. The analysis is also conducted in a conservative manner
containing margins in the calculation of mechanical analysis, iodine
inventory and iodine decontamination factor. Each of these
conservatisms will further decrease the consequences. Therefore, the
proposed changes do not significantly increase the probability or
consequences of any previously evaluated accident.
The proposed limits are used to establish operational conditions
where specific activities represent situations where significant
radioactive releases can be postulated. In addition, the changes to
operation are consistent with previous limits -- only allowing
increased flexibility after the radiological consequences are
assured to remain within accepted limits. Therefore, these
operational conditions are consistent with the design basis
analysis. The proposed changes do not introduce any new modes of
plant operation and do not involve physical modifications to the
plant. Therefore, the proposed changes do not create the possibility
of a new or different kind of accident from any previous analyzed.
The revised limits are used to establish operational conditions
where specific activities represent situations where significant
radioactive release can be postulated. These operational conditions
are consistent with the design basis analysis and are established
such that the radiological consequences are at or below the current
RBS licensing limit. Safety margins and analytical conservatisms
have been evaluated and are well understood. Conservative methods of
analysis are maintained through the use of accepted methodology and
benchmarking the proposed methods to previous analysis. Margins are
retained to ensure that the analysis adequately bounds all
postulated event scenarios. The proposed change only eliminates some
excess conservatism from the analysis.
EOI has implemented NUMARC 91-06 guidelines for shutdown
operations at RBS. Shutdown Operations Protection Plan and Primary-
Secondary Containment Integrity procedures presently include
guidance for closure of the containment hatch and other significant
opening in containment, in addition to the requirements contained in
the license and design basis. This additional protection will
enhance the ability to limit offsite effects.
Acceptance limits for the fuel handling accident are provided in
10CFR100 with additional guidance provided in NUREG 0800, Section
15.7.4 Excess margin is the difference between the postulated doses
and the corresponding licensing limit. In the
[[Page 47620]]
initial review of River Bend Station for operation (NUREG-0989, Section
15.7.4), the NRC accepted the design and analysis based on meeting
the guideline dose limits of 10CFR100 and SRP 15.7.4. The proposed
applicability continues to ensure that the whole-body and thyroid
doses at the exclusion area and low population zone boundaries, as
well as control room doses, are below the corresponding licensing
limit. These margins are unchanged; 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: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, DC 20005
NRC Project Director: William D. Beckner
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: June 20, 1995 (AEP:NRC:0692CX)
Description of amendment requests: The proposed amendments would
remove the requirements for fire protection systems from the licenses
and the Technical Specifications (T/S) in accordance with the
provisions and guidance of Generic Letters (GL) 86-10, ``Implementation
of Fire Protection Requirements,'' 88-12, ``Removal of Fire Protection
Requirements from Technical Specifications,'' and 93-07, Modification
of the Technical Specification Administrative Control Requirements for
Emergency and Security Plans.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
We have evaluated the proposed T/S changes and have determined
that the changes should involve no significant hazards consideration
based on the criteria established in 10 CFR 50.92(c). Operation of
CNP [Cook Nuclear Plant] in accordance with the proposed amendment
will not satisfy any of the following criteria.
(a) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are administrative in nature, in that it
moves the T/Ss portion of the Fire Protection Program from the T/Ss
to the UFSAR [Updated Final Safety Analysis Report] and the
implementing procedures. This is accomplished by referencing in the
UFSAR and the documents which address the Fire Protection Program in
greater detail. Thus, the proposed changes will not revise the
requirements for fire protection equipment operability, testing, or
inspection, but only moves the T/Ss portion of the Fire Protection
Program to implementing procedures.
As fire protection requirements are only being relocated
following the guidance of GLs 86-10, 88-12, and 93-07, the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(b) Create the possibility of a new or different kind of
accident from any previously analyzed.
The proposed changes do not involve any physical alteration of
plant configurations, changes to setpoints, or operating parameters.
[These] are administrative changes that retain the existing fire
protection requirements and relocate these requirements from the T/S
to the UFSAR; therefore, these changes do not create the possibility
of a new or different kind of accident.
(c) Involve a significant reduction in a margin of safety.
The proposed changes follow guidance contained in GLs 86-10, 88-
12, and 93-07 for incorporating the Fire Protection Program into the
UFSAR. A license condition will be implemented that will require
that no changes can be made to the Fire Protection Program that will
adversely affect the ability to achieve or maintain safe shutdown in
the event of a fire without prior NRC approval. 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 requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London,
Connecticut
Date of amendment request: August 23, 1995
Description of amendment request: The proposed amendment would
revise Technical Specifications Section 3.8.1.1 and the Bases for
Section 3/4.8. The proposed amendment would extend the Allowed Outage
Time (AOT) for an Emergency Diesel Generator (EDG) from 72 hours to 7
days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
... NNECO concludes that these changes do not involve a
significant hazards consideration since the proposed change
satisfies the criteria of 10 CFR 50.92(c). That is, the proposed
changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The EDGs supply backup power to the essential safety systems in
the event of a Loss of Normal (offsite) Power. EDGs are not accident
initiators. Therefore, this change does not involve an increase in
the probability of any accident previously evaluated.
Although the EDGs provide backup power to components that help
mitigate the consequences of accidents previously evaluated, the
extension in the AOT does not affect any of the assumptions used in
the deterministic evaluations of these accidents. Thus, this change
will not increase the consequences of any accident previously
analyzed.
The increase in the EDG AOT introduces the potential to increase
the risk to the public since a longer time window provides an
opportunity to perform additional preventive maintenance to the EDG
while the plant is on-line. However, the extended AOT, by itself,
does not necessarily increase risk. The increase in the risk depends
on the total time during which an EDG was out of service and the
other equipment that is concurrently out of service with the EDG.
The total risk increase due to the EDG being out of service will not
be significant since that risk increase is monitored and kept at
acceptable levels in accordance with the risk monitor program.
Based on the above, the proposal to extend the AOT for the EDGs
(Technical Specification 3.8.1) does not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change to extend the AOT for the EDGs (Technical
Specification 3.8.1) does not alter the physical design,
configuration, or method of operation of the plant. Therefore, the
proposal does not create the possibility of a new or different kind
of accident from any previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposed change to extend the AOT for the EDGs (Technical
Specification 3.8.1) do not affect the Limiting Conditions for
Operations or their bases. As a result, the deterministic analyses
performed to establish the margin of safety are unaffected. Thus,
the change does not involve a significant reduction in the margin of
safety.
[[Page 47621]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London,
Connecticut
Date of amendment request: August 23, 1995
Description of amendment request: The proposed amendment would
extend the Allowed Outage Time (AOT) for an inoperable Safety Injection
Tank (SIT) from 1 hour to 24 hours, unless the SIT is inoperable due to
either boron concentration not within its limits or an inoperable level
or pressure instrument. For these two special cases, the proposed
change extends the AOT for an inoperable SIT to 72 hours. In addition,
the proposed amendment clarifies the completion times and conditions
for action statements and the criteria for surveillance requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
... NNECO concludes that these changes do not involve a
significant hazards consideration since the proposed change
satisfies the criteria in 10 CFR 50.92(c). That is, the proposed
changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The Safety Injection Tanks (SITs) are passive components in the
Emergency Core Cooling System that mitigate the consequences of a
Loss of Coolant Accident (LOCA). As such, the SITs are not accident
initiators. Therefore, this change does not involve an increase in
the probability of any accident previously evaluated.
The increase in the AOT has the potential to increase the risk
if it becomes necessary to stay on-line longer than one (1) hour
with an inoperable SIT. However, the estimated risk impact is
negligible.
The SITs inject borated water into the reactor vessel (via the
cold legs) during the blowdown phase of a large break LOCA. The
introduction of the inventory of borated water from all four (4)
SITs is needed to ensure adequate reflooding of the core (i.e.,
minimize core damage) until the Engineered Safety Feature (ESF)
pumps can provide adequate core cooling. The SITs also provide
makeup water for the Reactor Coolant System (RCS) for smaller break
LOCAs. The extension of the AOT does not affect any of the
assumptions used in the deterministic evaluations of these
accidents. Thus, this change will not increase the consequences of
any accident previously evaluated.
The increased AOT extension to 72 hours, based solely on
instrumentation (level and pressure) malfunction, also does not
involve a significant increase in the consequences of an accident
previously evaluated as endorsed by the NRC in NUREG-1366,
``Improvements to Technical Specifications Surveillance
Requirements.''
The modification to the completion times and the modification of
the Surveillance Requirements for volumetric changes in the SIT as a
result of addition from the Refueling Water Storage Tank (RWST) also
do not involve a significant increase in the consequences of any
accident previously evaluated by the NRC in NUREG-1432, ``Standard
Technical Specifications for Combustion Engineering Plants.''
Based on the above, the proposed changes to extend the AOT for
an inoperable SIT, clarify action statements, and modify the
criteria for surveillance requirements, do not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes to extend the AOT for an inoperable SIT,
clarify action statements, and modify the criteria for surveillance
requirements, do not alter the physical design, configuration, or
method of operation of the plant. Therefore, the proposal does not
create the possibility of a new or different kind of accident from
any previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposed changes to extend the AOT for an inoperable SIT,
clarify action statements, and modify the criteria for surveillance
requirements, do not affect the Limiting Conditions for Operations
(LCOs) of the SITs or the bases of the LCOs. As a result, the
deterministic analyses performed to establish the margin of safety
are unaffected. Thus, the change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: May 4, 1995.
Description of amendment requests: The proposed amendments would
revise the pressurizer and main steam safety valve lift setting
tolerances from plus or minus 1% to plus or minus 3%, revise the Safety
Limit curves and revise the Technical Specification Section 2 to
conform to Standard Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated
The proposed changes increase the ``as-found'' setpoint
tolerances for the Pressurizer Safety Valves and Main Steam Safety
Valves from [plus or minus] 1% to [plus or minus] 3%. The proposed
changes do not involve any hardware modifications to plant
structures, systems, or components. Analyses have determined that
the proposed changes do not significantly affect the structural
integrity of either the Reactor Coolant System or the Main Steam
system.
The proposed setpoint tolerance of [plus or minus] 3% was
included in the assumptions for the performance of the reload safety
evaluations for the current fuel cycles, PI1-17 and PI2-16, and
subsequent Prairie Island fuel cycle analyses. These analyses
concluded that the minimum acceptable DNBR [departure from nucleate
boiling ratio] is maintained, over-pressure protection is
maintained, LOCA [loss-of-coolant accident] acceptance criteria are
met and offsite dose limits are not exceeded. These changes are
consistent with the guidance provided by Section III and XI of the
ASME [American Society of Mechanical Engineers] Code and Standard
Technical Specifications.
The proposed change to Technical Specification Figure TS.2.1-1
does not affect any existing accident analyses. This revision
ensures that the design bases and safety limits are accurately and
appropriately reflected in the Technical Specifications and will
ensure that plant operations are properly evaluated for DNBR
encroachment.
Therefore, the probability or consequences of an accident
previously evaluated are not affected by any of the proposed
amendments.
2. The proposed amendment will not create the possibility of a
new of different
[[Page 47622]]
kind of accident from any accident previously analyzed The lift
setpoint the Pressurizer Safety Valves and Main Steam Safety Valves
will be restored to [plus or minus] 1% following testing, thus the
``as-left'' setpoint tolerance for the Pressurizer Safety Valves and
Main Steam Safety Valves are unchanged. Evaluations of plant normal
operation, transient and accident conditions have been performed
assuming these safety valve lift settings are [plus or minus] 3% of
the nominal setpoint and demonstrated that new or different kinds of
accidents are not created by the proposed changes.
The proposed changes to Technical Specification Figure TS.2.1-1
do not affect the design, function or operation of any Prairie
Island structures, systems or components. The curves show the loci
of points of reactor core differential temperature (an indication of
thermal power), Reactor Coolant System pressure, and average
temperature for which the minimum DNBR is not less than the safety
analysis limit, that fuel centerline temperature remains below
melting, that the average enthalpy in the hot leg is less than or
equal to enthalpy of saturated liquid, or that the exit quality is
within the limits defined by the applicable DNBR correlation. There
are no new failure modes introduced by the proposed changes to the
Figure. The changes conservatively adjust Figure TS.2.1-1 to current
plant conditions and ensure that the design is accurately reflected
and that the plant is operated in accordance with its design bases.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be created
[by] these amendments.
3. The proposed amendment will not involve a significant
reduction in the margin of safety
The proposed changes to the safety valve lift setting tolerances
are consistent with the guidance provided by Section III and XI of
the ASME Code and Standard Technical Specifications. Analyses have
demonstrated these valves will continue to perform their function of
protecting their respective system from over-pressurization under
all postulated transients and accidents. The changed setting
tolerances do not cause a reduction in any other safety margin such
as DNBR. SAFETY LIMIT curves are provided to define minimum
allowable safety margin for plant steady state operation, normal
operational transients and anticipated operational occurrences. The
SAFETY LIMITs represent a design requirement for establishment of
many of the RPS [reactor protection system] trip setpoints which
prevent reactor conditions from approaching the SAFETY LIMITs. The
proposed revision of the SAFETY LIMIT curves provide the minimum
safety margins with somewhat more conservatism than previously
included. No RPS trips setpoints are changed.
Therefore, a significant reduction in the margin of safety would
not be involved with these amendments.
Based on the evaluation described above, and pursuant to 10 CFR
Part 50, Section 50.91, Northern States Power Company has determined
that operation [of] the Prairie Island Nuclear Generating Plant in
accordance with the proposed license amendment request does not
involve any significant hazards considerations as defined by Nuclear
Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis, MN
55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendment would
revise the 250 volt DC [direct current] profiles in Technical
Specifications Surveillance 4.8.2.1 (d) (2c) to reflect the new load
profile calculations.
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.
[Final Safety Analysis Report] FSAR Section 8.3 states that the
station batteries have sufficient capacity without the charger to
independently supply the required loads for four hours. The
Technical Specifications require that the batteries be surveilled to
dummy loads which are greater than the design loads. The load
profiles for the 250 VDC batteries were recalculated using discrete
increments of time when the loads would be in use for each of five
design basis events. The Technical Specification load profiles are a
composite of the worst case loads for the events plus margin. The
required ampere-hours for each battery using the new load profiles
is less than the ampere-hours required using the existing load
profiles. Therefore, since the load profiles envelop the actual
loads on the batteries, the change to the 250 VDC battery load
profiles does not involve a significant increase in the probability
or consequence of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
As stated above, the 250 VDC batteries have sufficient capacity
to power the actual battery loads thus enabling them to perform
their intended function. Any postulated accident resulting from this
change is bounded by previous analysis. Therefore, the change to the
250 VDC battery load profiles does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The Class 1E 250 VDC batteries are required to have sufficient
capacity and capability to ensure sufficient power is available to
supply the safety related equipment for (1) the safe shutdown of the
facilities and (2) the mitigation and control of accident conditions
within the facilities. The proposed load profiles envelope the worst
case loads plus margin.
The ampere-hours removed from the Class 1E 250 VDC batteries are
less for the proposed load profiles than the existing load profiles.
The ampere-hours available in the batteries after the batteries have
supplies[d] the emergency loads for 4 hours are: [See table in
subject application].
* * * * * * *
Engineering calculation shows that the Class 1E 250 VDC
batteries maintain at least 210 VDC at the Class 1E 250 VDC MCCs
while supplying the proposed loads, corrected for temperature and
aging. Since the Class 1E 250 VDC circuits are designed to operate
properly with a minimum of 210 VDC at the Class 1E MCCs, all the
Class 1E emergency equipment supplied from the Class 1E batteries
have sufficient voltage to operate for 4 hours after the loss of ac
power.
The Class 1E 250 VDC batteries and Class 1E 250 VDC battery
chargers have been sized using the proposed load profiles. The
Engineering calculation shows that the 120 cell, 12 positive plates
per cell battery banks are sufficient to supply the proposed load
profiles, corrected for temperature and aging. The same calculation
also shows that the Class 1E 250 VDC battery chargers have
sufficient capacity to re-charge the batteries from the proposed
emergency discharged conditions to the fully charged condition in 12
hours while continuing to supply the plant normal continuous loads.
Base upon the above discussion, the proposed changes to the
Technical Specification load profiles do not reduce the margin of
safety as defined in the Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and
[[Page 47623]]
Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket No. 50-388,
Susquehanna Steam Electric Station, Unit 2, Luzerne County,
Pennsylvania
Date of amendment request: August 11, 1995
Description of amendment request: The proposed amendment would
revise Susquehanna Unit 2 Technical Specification Table 3.3.7.5-1 as
follows:a.
Revise Item 13, Required Number of Channels from 1 to 2;b.
Revise Item 13, Minimum Channel Operable from 0 to 1;c.
Delete Footnote .
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:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Reestablishing the channel operability values in Item
13 of Technical Specification Table 3.3.7.5-1, and deleting
footnote , has no impact on the
probability or consequences of an accident previously evaluated. The
proposed change in the channel operability values is a return to the
values which were reviewed as part of the licensing basis.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Reestablishing the channel operability values in Item
13 of Technical Specification Table 3.3.7.5-1, and deleting
footnote , does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The change in the channel operability values
increases the required number of channels available for accident
monitoring. There is no correlation between increasing the number of
neutron flux accident monitoring channels available and the creation
of accident scenarios.
III. This change does not involve a significant reduction in a
margin of safety.
Reestablishing the channel operability values in Item
13 of Technical Specification Table 3.3.7.5-1, and deleting
footnote , does not involve a reduction
in a margin of safety. The proposed change increases the number of
required channels from current levels, and restores the values to
those which have historically been required. At the present time,
the number of required channels is being administratively controlled
at the proposed levels to ensure conservatism in operability.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: May 12, 1995
Description of amendment request: The proposed change would extend
the surveillance test intervals for the emergency service water (ESW)
system to support 24 month operating cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92 since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes increase the interval between ESW system
surveillance tests. These changes are consistent with the guidance
provided in Generic Letter 91-04. These changes do not involve any
physical changes to the plant, nor do they alter the typical way the
ESW system functions. On-line testing will continue to assure
equipment availability. The type of testing and the corrective
actions required if the subject ESW surveillances fail remain the
same. As such, the proposed changes create no new impacts on
accidents previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes increase the interval between ESW system
surveillance tests. These changes are consistent with the guidance
provided in Generic Letter 91-04. The proposed changes do not change
the ability of the ESW system to provide heat removal for the ECCS
[emergency core cooling system] components and other equipment
essential to reactor shutdown. Past equipment performance and on-
line testing indicate the longer test intervals will not degrade ESW
equipment. No changes are proposed to the type of testing performed,
only to the length of the surveillance interval. The proposed
changes do not modify the design or operation of plant equipment,
therefore, no new or different failure modes are introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. involve a significant reduction in a margin of safety.
The proposed changes increase the interval between ESW system
surveillance tests. These changes are consistent with the guidance
provided in Generic Letter 91-04. The proposed changes do not alter
the configuration of the ESW system nor change the manner in which
the ESW equipment functions. Past equipment performance and on-line
testing indicate the longer test intervals will not degrade ESW
equipment. Operation of the plant remains unchanged by the proposed
changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, NY
13126
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, NY 10019
NRC Project Director: Ledyard B. Marsh
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: June 15, 1995
Description of amendment request: The proposed change would extend
the surveillance test intervals for the control rod system to support
24 month operating cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
11. involve a significant increase in the probability or consequences
of an accident previously evaluated.
[[Page 47624]]
The proposed changes increase the interval between control rod
system surveillance tests. These changes are consistent with the
guidance provided in Generic Letter 91-04. These changes do not
involve any physical changes to the plant, nor do they alter the way
the control rod system functions. The type of testing and the
corrective actions required if the subject control rod surveillances
fail remain the same. As such, the proposed changes create no new
impacts on accidents previously evaluated.
The reactivity margin - core loading test can be safely extended
to accommodate the 24 month operating cycle. The calculation of
reactivity margin takes into account the longer operating cycle.
The control rod scram time test can be safely extended to
accommodate a 24 month operating cycle. Operating experience has
indicated that control rod scram times do not significantly change
over an operating cycle. Additional on-line testing provides
adequate assurance of equipment operability.
The SDIV [Scram Discharge Instrument Volume] vent and drain
valve operability test can be safely extended to accommodate a 24
month operating cycle. Evaluation of past surveillance performance
and additional on-line testing assure valve operability. The
operability of the mode switch and the reset switch is demonstrated
during shutdowns.
Therefore, the proposed changes do not involve a significant
increase in the probability and do not change the consequences of an
accident previously evaluated.
2. create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes increase the interval between control rod
system surveillance tests. These changes are consistent with the
guidance provided in Generic Letter 91-04. The proposed changes do
not change the ability of the control rod system to provide rapid
reactivity control in order that no fuel damage results from any
abnormal operating transient. Past equipment performance and on-line
testing indicate the longer test intervals will not degrade control
equipment. No changes are proposed to the type of testing performed,
only to the surveillance interval length. The proposed changes do
not modify the design or operation of plant equipment, therefore, no
new or different failure modes are introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. involve a significant reduction in a margin of safety.
The proposed changes increase the interval between control rod
system surveillance tests. These changes are consistent with the
guidance provided in Generic Letter 91-04. The proposed changes do
not alter the configuration of the control rod system nor change the
manner in which the control rod system functions. Past equipment
performance and on-line testing indicate the longer test intervals
will not degrade control rod equipment. Operation of the plant
remains unchanged by the proposed changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, NY
13126
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, NY 10019
NRC Project Director: Ledyard B. Marsh
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: July 21, 1995
Description of amendment request: The proposed changes would
replace the title-specific list of members on the Plant Operating
Review Committee (PORC) with a more general statement of membership
requirements, similar to that used to define Safety Review Committee
membership; expand the scope of disciplines represented on the PORC to
include Nuclear Licensing and Quality Assurance; change several
management position titles; and, make several editorial corrections to
the Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Replacing the title specific list of PORC members with a
statement of membership requirements for the committee does not
reduce the effectiveness of the committee to advise the Resident
Manager (Site Executive Officer) on matters regarding nuclear
safety.
The proposed title changes for the Chief Nuclear Officer, Site
Executive Officer, Shift Manager, and Control Room Supervisor are
changes in title only and do not affect the responsibilities,
authority, qualification requirements, or reporting relationships of
these positions.
The change proposed for Specification 6.12 is administrative in
nature, reflecting a change previously approved elsewhere in
Technical Specifications.
The Radiological and Environmental Services Manager title change
proposed for Specification 6.11(A)2 is administrative in nature,
reflecting a change previously approved elsewhere in Technical
Specifications.
The remainder of proposed changes correct grammar or improve
consistency in Technical Specification formatting and do not affect
the meaning or intent of the specifications involved.
Operation of the James A. FitzPatrick Nuclear Power Plant in
accordance with the proposed amendment would not involve a
significant hazards consideration as defined in 10 CFR 50.92. The
changes are administrative in nature and would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. create the possibility of a new or different kind of accident
from those previously evaluated, or
3. involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, NY 10019
NRC Project Director: Ledyard B. Marsh
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: July 27, 1995
Description of amendment request: The proposed change to the
Technical Specifications (TS) would incorporate updated pressure vs.
temperature operating limit curves contained in TS Figure 3.4.6.1-1 and
revise TS Surveillance Requirement 4.4.6.1.3 based on implementation of
Regulatory Guide 1.99, Rev. 2 in accordance with Generic Letter 88-11.
The changes are a result of data obtained from the first set of
specimen capsules removed during Refueling Outage 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. Will not involve a significant increase in the probability or
consequences of an accident [...] previously evaluated.
The proposed changes assure that the existing safety limits are
not exceeded due to changing Reactor Vessel conditions. These
changes reflect the latest material testing
[[Page 47625]]
results in accordance with 10CFR50, Appendix G. The proposed changes to
the pressure and temperature limits do not increase the probability
of nonductile failures. The proposed changes to the surveillance
requirement and the associated changes to the Bases to include a
commitment to the methodology of Regulatory Guide 1.99, Rev. 2
ensures that the most limiting Reactor Vessel material is used in
the determination of the pressure-temperature operating limits.
Therefore, it may be concluded that the proposed changes do not
involve a significant increase in the probability or consequences of
an accident or malfunction of equipment important to safety
previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
No physical plant modifications or new operating configurations
result from these changes. These changes do not adversely affect the
design or operation of any system or component important to safety,
rather they establish limits to assure that operations remain within
acceptable safety boundaries.
Therefore, the possibility of a new or different kind of
accident from any previously evaluated will not be created.
3. Will not involve a significant reduction in a margin of
safety. Analysis of the capsule specimens has concluded that the
Reactor Vessel has sufficient fracture toughness for continued safe
operation, provided that operation remains within acceptable
pressure-temperature limits. The revised pressure-temperature curves
define these acceptable pressure-temperature limits during plant
operation. The proposed changes maintain the existing margins of
safety by modifying the operating limits based on the most limiting
of the actual reference temperature shifts. This new limit
considered analytical results of the capsule specimens, or a
predicted shift considering the most limiting pressure vessel
material. Changes to the Surveillance Requirement criteria and the
associated Bases to include a commitment to the methodology
contained in Regulatory Guide 1.99, Rev. 2 will ensure that the most
limiting plate or beltline weld material will be utilized in the
determination of the pressure-temperature limits.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
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 amendment requests: August 1, 1995
Description of amendment requests: The amendment request proposes
to change Technical Specification (TS) 3/4.3.2, Table 3.3-3,
``Engineered Safety Features Actuation System Instrumentation.'' TS 3/
4.3.2 includes the requirements for the minimum number of toxic gas
isolation system (TGIS) trains operable. The TS change request is to
extend the allowed TGIS outage times during replacement of TGIS
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Toxic Gas Isolation System (TGIS) is designed to monitor and
mitigate the effects of toxic gas releases on control room
habitability. TGIS unavailability is not a precursor to any accident
previously evaluated in Chapter 15 of the San Onofre Updated Final
Safety Analysis Report (UFSAR). A risk assessment of the TGIS
instrumentation replacement activity was performed and found that
the likelihood of a loss of control room habitability beyond that
permitted by the Technical Specifications (TS) will not exceed 1E-6
over the duration of this TS change. In addition, a loss of control
room habitability does not necessarily lead to an accident or core
damage event. However, if a loss of control room habitability was
conservatively assumed to lead to a core damage event, this increase
in risk would still not constitute a significant increase in the
consequences or probability of any accident previously evaluated
since the increase is less than 3% of the average annual core damage
risk from internal events as reported in the San Onofre Individual
Plant Examination. Therefore, operation of the facility in
accordance with this proposed change does not involve a significant
increase in the probability or consequences of any 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.
This change extends the allowed outage times of the TGIS system.
The change does not affect the design or operation of any other
plant systems. An increase in TGIS unavailability is not a precursor
to any accident previously evaluated in Chapter 15 of the San Onofre
UFSAR. Therefore, operation of the facility in accordance with this
proposed change 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.
During replacement of TGIS instrumentation a single channel of
TGIS will be maintained operable except during periods when
construction activity may result in spurious TGIS alarms. During
these periods the control room will normally be isolated except for
brief periods when the control room will be open to allow for air
exchange or to allow for CREACUS equipment repair. These periods,
when the control room is open without a TGIS channel available, will
not exceed 54 hours during the entire period when this change is in
effect. Operation with control room ventilation in the normal mode
with a single channel of TGIS operable for 44 days and no TGIS
channel available for up to 54 hours has been analyzed, and results
in an increase in the probability of a loss of control room
habitability which does not exceed 1E-6 over the duration of this TS
change. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, CA 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, CA 91770
NRC Project Director: William H. Bateman
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
[[Page 47626]]
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: October 24, 1994, as
supplemented July 21, 1995. The July 21, 1995, letter provides
clarification information and did not change the scope of the October
24, 1994, letter, or the initial no significant hazards consideration
determination.
Brief description of amendment: The proposed amendment would revise
the TS to allow the relocation of TS 3/4.3.7.12, Area Temperature
Monitoring; and the associated Bases in the TS to licensee-controlled
documents.
Date of issuance: August 28, 1995
Effective date: August 28, 1995
Amendment No.: 62
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 23, 1994 (59
FR 60379) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 28, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, NC 27605
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: May 18, 1995, as supplemented
May 31, 1995
Brief description of amendments: The amendments revise the
frequency for conducting the Catawba Unit 2 Integrated Leak Rate Test
(ILRT) from a nominal frequency of once per 40 months to less than or
equal to 70 months. This also involves the granting of an exemption
from the requirements of 10 CFR Part 50, Appendix J, which is addressed
by separate correspondence.
Date of issuance: August 18, 1995
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: 133 and 127
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32362) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 18, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, SC 29730
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley
Power Station, Unit 2, Shippingport, Pennsylvania
Date of application for amendment: April 26, 1995
Brief description of amendment: This amendment adds a requirement
to Technical Specification (TS) 4.5.2.a to periodically verify that the
High Head Safety Injection (HHSI) pump minimum flow valve, 2CHS*MOV373,
is maintained open during plant operation in Modes 1, 2, and 3. Valve
2CHS*MOV373, must be maintained open to provide a minimum flowpath for
the HHSI pumps thereby minimizing the likelihood of HHSI pump damage
due to pump operation with insufficient flow. The amendment allows
flexibility for local verification of valve position or flow indication
if the control room indication is not available. Several editorial
changes to TS 3/4.5.2 are also being made to provide consistent format
with other TSs.
Date of issuance: August 25, 1995
Effective date: August 25, 1995
Amendment No.: 73
Facility Operating License No. NPF-73: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29874). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 25, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket No.
50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County,
Georgia
Date of application for amendment: April 14, 1995, as supplemented
by letters dated June 22 and July 18, 1995
Brief description of amendment: The amendment eliminates response
time testing (RTT) requirements for selected sensors and specific loop
instrumentations for (1) the Reactor Protection System (RPS), (2) the
Isolation System, and (3) the Emergency Core Cooling System (ECCS). In
addition, the Note for Surveillance Requirement 3.3.6.1.7, which reads:
``Radiation detectors may be excluded,'' is being removed since RTT is
not required for any radiation detector that provides a primary
containment isolation signal as indicated in Table 3.3.6.1-1 of the TS.
Date of issuance: August 23, 1995
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance
Amendment No.: 137
Facility Operating License No. NPF-5: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35076) The June 22 and July 18, 1995, letters provided clarifying
information that did not change the scope of the April 14, 1995,
application and initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 23, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, GA 31513
[[Page 47627]]
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: January 3, 1995, as
supplemented by letters dated June 14 and July 6, 1995.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) with editorial changes to the Action
Statements of TS 3.8.1.1 and 3.8.1.2 in order to reflect the
availability of a third offsite ac electrical source. Technical
Specification 4.8.1.1.1 is clarified to specify that the offsite ac
circuits connected to the onsite Class 1E distribution system are
required to be verified OPERABLE. A footnote is added to TS 3.8.3.1 to
allow the connection of the third offsite ac source to the onsite
busses.
Date of issuance: August 29, 1995
Effective date: As of the date of issuance to be implemented
within 30 days from the date of issuance
Amendment Nos.: 90 and 68
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6301) The June 14 and July 6, 1995, letters provided clarifying
information that did not change the scope of the January 3, 1995,
application and initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 29, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, GA 30830
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: May 23, 1995
Brief description of amendments: The amendments revise the column
format for the Reactor Protection System and Engineered Safety Feature
Actuation System Setpoints
Date of issuance: August 24, 1995
Effective date: August 24, 1995
Amendment Nos.: 176 and 170Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32364) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 24, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, FL 33199
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: December 20, 1993, as
supplemented July 19, 1994, and February 28, 1995.
Brief description of amendments: The amendments revise the
surveillance requirements and load profiles for A, B, and N Train
batteries.
Date of issuance: August 22, 1995
Effective date: August 22, 1995
Amendment Nos.: 198 and 183
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4939) and June 6, 1995 (60 FR 29879) The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
August 22, 1995.No significant hazards consideration comments received:
No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: May 25, 1995, and supplemented
June 30, 1995
Brief description of amendments: The amendments allow fuel
reconstitution when analyzed in accordance with NRC-approved
methodologies. The amendments are line item improvements based on NRC
Generic Letter 90-02, ``Alternative Requirements for Fuel Assemblies in
Design Features Section of Technical Specifications,'' supplement 1.
Date of issuance: August 22, 1995
Effective date: August 22, 1995
Amendment Nos.: 199 and 184
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35081) The June 30, 1995, supplement provided a minor revision to the
proposed Technical Specification pages which was within the scope of
the original application and did not change the staff's initial
proposed no significant hazards considerations determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 22, 1995.No significant hazards
consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: February 14, 1995
Brief description of amendment: This amendment makes the following
administrative changes to the Maine Yankee (MY) Technical
Specifications (TS):
a. Removes responsibility for audits of the emergency and security
plans--including their implementing procedures--from the TS and assigns
that responsibility to the emergency and security plans,
b. Assigns review responsibility for significant, accidental,
unplanned, or uncontrolled radioactive releases to the Nuclear Safety
Audit and Review (NSAR) Committee,
c. Assigns additional reporting requirements to the NSAR Committee,
and
d. Provides the President of MY with the authority to initiate an
audit of any area of facility operation.
Date of issuance: August 22, 1995
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 152
Facility Operating License No. DPR-36: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16191) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 22, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: May 18, 1995
Brief description of amendment: The amendment revises the minimum
temperature at which the reactor vessel head bolting studs are allowed
to be
[[Page 47628]]
placed under tension. In addition, the amendment revises the minimum
reactor vessel metal temperature during core critical operation,
revises the minimum reactor vessel metal temperature for pressure
tests, makes editorial changes, and revises the Bases for the
applicable section.
Date of issuance: August 23, 1995
Effective date: As of the date of issuance to be implemented
immediately.
Amendment No.: 85
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32369) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 23, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: June 15, 1995
Brief description of amendment: The amendment changes the
definition for an alteration of the reactor core to one that is
consistent with the intent of the improved standard technical
specifications. The amendment also makes administrative changes to
several technical specification pages.
Date of issuance: August 28, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 86
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37097) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 28, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: April 28, 1995, as supplemented
August 2, 1995.
Brief description of amendment: The amendment changes Technical
Specification (TS) Sections 3.7.5, 4.7.5, and 3/4.7.5, to permit
Millstone Unit 3 to remain in operation with the average ultimate heat
sink water temperature greater than 75* F (but less than or equal to
77* F) for a period of 12 hours.
Date of issuance: August 28, 1995
Effective date: As of the date of issuance.
Amendment No.: 119
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29881). The information in the licensee's submittal of August 2, 1995,
did not require a change to the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated August 28, 1995.No
significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: June 29, 1995
Brief description of amendments: The amendments revise the combined
Technical Specifications (TS) for Diablo Canyon Nuclear Power Plant,
Unit Nos. 1 and 2 (DCPP) to add Mode 1 applicability to TS 3/4.4.2.2,
``Safety Valves - Operating,'' and changes the low-temperature
overpressure protection (LTOP) system enable temperature for Mode 4
applicability from 323 degrees F to 270 degrees F in TS 3/4.3.2.1,
``Safety Valves - Shutdown.''
Date of issuance: August 23, 1995
Effective date: August 23, 1995
Amendment Nos.: Unit 1 - Amendment No. 107; Unit 2 - Amendment No.
106
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37098) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 23, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, CA 93407
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: February 1, 1995, as
supplemented by letter dated June 20, 1995
Brief description of amendments: The requested changes would modify
the applicable operational conditions for the secondary containment
isolation radiation monitors located on the refueling floor and for the
monitor located in the railroad access shaft.
Date of issuance: August 24, 1995
Effective date: Both units, as of the date of issuance and is to be
implemented within 30 days
Amendment Nos.: 152 and 122
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16192). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 24, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: August 31, 1994, as
supplemented by letters dated May 11, and July 3, 1995
Brief description of amendments: This amendment revises the
Technical Specifications to permit the relocation of the Turbine
Overspeed Protection System to the Updated Final Safety Analysis Report
and Controlled Plant Procedures.
Date of issuance: August 24, 1995
Effective date: August 24, 1995
Amendment Nos.: 100 and 64
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55884) The supplemental letters do not
[[Page 47629]]
change the initial no significant hazards consideration determination
nor the initial Federal Register notice. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
August 24, 1995.No significant hazards consideration comments received:
No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: February 22, 1995
Brief description of amendments: The amendments revise the
Technical Specifications Surveillance Requirements to clarify the
Emergency Diesel Generator acceptable steady state voltage range.
Date of issuance: August 24, 1995
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 101 and 65
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20525) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 24, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: January 13, 1995
Brief description of amendment: The amendment revised the
Administrative Controls Section (6.0) of the Technical Specifications
for Hope Creek Generating Station to reflect organizational changes and
resultant management title changes.
Date of issuance: August 22, 1995
Effective date: August 22, 1995
Amendment No.: 77
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32371) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 22, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: December 23, 1994
Brief description of amendments: The amendments to the Technical
Specifications revise the surveillance requirement to perform a visual
inspection of containment areas affected by containment entry when
containment integrity is established. They are consistent with Item 7.5
of Generic Letter 93-05, ``Line-Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation.''
Date of issuance: August 24, 1995
Effective date: As of the date of issuance, to be implementd within
60 days.
Amendment Nos.: 174 and 155
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6308) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 24, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: September 16, 1994
Brief description of amendments: These amendments revise Technical
Specification (TS) 3/4.2.1, ``Linear Heat Rate.'' The linear heat rate
(LHR) limit for steady state operation is revised from 13.9 kw/ft to
13.0 kw/ft. The Bases for TS 3/4.2.1, ``Linear Heat Rate,'' is also
being revised to reflect the new value.
Date of issuance: August 23, 1995
Effective date: August 23, 1995, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2 - Amendment No. 124; Unit 3 - Amendment No.
113
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55892) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 23, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, CA 92713
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: May 3, 1995
Brief description of amendments: The amendments delay
implementation of Amendment Nos. 182 and 174 until implementation
problems are addressed. These changes revise the setpoints and time
delays for the auxiliary feedwater loss of power and the 6.9 kv
shutdown board loss of voltage and degraded voltage instrumentation.
Date of issuance: August 22, 1995
Effective date: August 22, 1995
Amendment Nos.: 207 and 197
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27343) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 22, 1995.No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, TN 37402
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 6, 1995 (TS 94-18)
Brief description of amendments: The amendments revise Surveillance
Requirement 4.0.5 by replacing the current Inservice Inspection program
and the Inservice Testing program requirements with the requirements
stated in the Standard Technical Specifications (NUREG-1431). The
amendments also delete Technical Specification 3/4.4.10, ``Structural
Integrity ASME Code Class 1, 2 and 3 Components,'' and its related
Bases information.
Date of issuance: August 22, 1995
Effective date: August 22, 1995
Amendment Nos.: 208 and 198
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20528)
[[Page 47630]]
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated August 22, 1995.No significant hazards
consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, TN 37402
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: April 17, 1995, as supplemented
on June 30, 1995
Brief description of amendment: The amendment revises Technical
Specifications Technical Specification 2.2.1, Table 2.2-1. The changes
address reducing repeated alarms and partial reactor trips by revising
the Overpower Delta-T setpoint function.
Date of issuance: August 21, 1995
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 102
Facility Operating License No. NPF-30. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24922). The June 30, 1995, letter provided supplemental 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 August 21, 1995. No
significant hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, MO 65251
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: September 2, 1992
Brief description of amendment: The amendment revises the required
signal-to-noise ratio for the source range monitors, as recommended by
General Electric.
Date of issuance: August 23, 1995
Effective date: August 23, 1995
Amendment No.: 140
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37101) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 23, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, WA 99352
Dated at Rockville, Maryland, this 6th day of September 1995.
For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV Office of Nuclear
Reactor Regulation.
[Doc. 95-22616 Filed 9-12-95; 8:45 am]
BILLING CODE 7590-01-F
