# Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** March 1, 1995
- **Citation:** 60 FR 11125

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Pubic Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission on 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
[[Page 11126]] 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 February 3, 1995, through February 16, 1995.
The last biweekly notice was published on February 15, 1995 (60 FR
8739).

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 March 31, 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
shale 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 [[Page 11127]] significant hazards consideration, the Commission may
issue the amendment and make it immediately effective, notwithstanding
the request for a hearing. Any hearing held would take place after
issuance of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of 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 section, 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: December 7, 1994.
Description of amendment requests: The proposed amendment would
revise the capacity of the ultimate heat sink (UHS) as described in the
bases of Technical Specification 3/4.7.5, ``Ultimate Heat Sink,'' from
providing a 27-day cooling water supply to providing a 26-day cooling
water supply. In addition, the reference to Regulatory Guide 1.27 in
the bases of this TS would also be revised to reference the January
1976 revision rather than the March 1974 revision.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:

Standard 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The Essential spray pond system and the UHS do not initiate any
accidents in Chapters 6 or 15 of the UFSAR [Updated Final Safety
Analysis Report]. The justification and basis for the time that the
UHS is available is not changed and continues to be consistent with
the guidance in Regulatory Guide 1.27. The existing Technical
Specification requirements and those components to which they apply
are not altered by this Technical Specification amendment.
Therefore, the change to the bases for Technical Specification 3/
4.7.5 does not increase the probability of occurrence or the
consequences of any previously evaluated accident.
Standard 2--Does the proposed change create the possibility of a
new or different kind of accident from any accident previously
evaluated?
The requirements for Technical Specification 3/4.7.5 are not
changed. This amendment has no impact on plant maintenance, testing,
shutdown equipment, or component qualification. Therefore, the
possibility of a new or different kind of accident is not created by
this amendment.
Standard 3--Does the proposed change involve a significant
reduction in a margin of safety?
The change to the bases for Technical Specification 3/4.7.5 does
not significantly alter existing Technical Specification
requirements or those coponments to which they apply. The
justification and basis for the time that the UHS is available
without makeup is not changed and continues to be consistent with
the guidance in Regulatory Guide 1.27. Regulatory Guide 1.27 states
that ``A capacity less than 30 days may be acceptable if it can be
demonstrated that replenishment can be effected to ensure that
continuous capability of the sink to perform its safety functions,
taking into account the availability of replenishment equipment and
limitations that may be imposed on ``freedom of movement'' following
an accident.'' This change does not effect the continuous capability
of the UHS to perform its safety function of providing decay heat
removal capability following an accident. The change updates the
design basis of the UHS using more realistic conditions based on
plant experience. Therefore, the change in the capacity of the UHS
without makeup from 27 days to 26 days will not involve a
significant reduction in margin of safety for the ultimate heat
sink.

The NRC staff has reviewed the licensees' analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Attorney for licensees: Nancy C. Loftin, Esq., Corporation
Secretary and Counsel, Arizona Public Service Company, P.O. Box 53999,
Mail Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: Theodore R. Quay.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: June 18, 1992, as supplemented December
8, 1992, and revised February 3, 1995.
Description of amendment request: The proposed Technical
Specification (TS) amendment adds limiting conditions of operation and
surveillance requirements for the pressurizer power-operated relief
valves (PORVs) and their associated block valves whenever average
temperature (Tavg) is above 350 degrees F or the reactor is critical.
Specifications have also been added for low-temperature overpressure
protection whenever Tavg is less than 350 degrees F and the reactor
coolant system is not vented to the containment. The February 3, 1995,
revision made editorial changes to previous TS pages and made changes
to conform with an additional provision of the guidance for
surveillance testing of the block valves associated with the
pressurizer PORVs. In addition, the licensee has requested an editorial
change to TS page 3.1.-11 to revise the references to two figures that
have been superseded.
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 requested revision does not involve a significant
increase in the probability or consequences of an accident
previously [[Page 11128]] evaluated. The proposed revision to our
previous Technical Specification (TS) change request dated June 18,
1992, would help assure the availability of the block valves for
accident mitigation. The availability of the block valves for
accident mitigation has been found to outweigh any negative safety
consequences associated with full cycle testing of a block valve
isolating a pressurizer power-operated relief valves (PORV) with
``excessive'' seat leakage. There would be no significant increase
in the probability or consequences of an accident previously
evaluated since this event is fully bounded by the failing open of a
single pressurizer code safety relief valve event which is analyzed
in Chapter 15 of the Updated Final Safety Analysis Report.
Accordingly, the requested revision will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The requested revision to our previous TS change request does
not create the possibility of a new or different kind of accident
from any accident previously evaluated. Periodic testing of the
block valves in accordance with the requested revision is only
intended to assure the functioning and capability of the block
valves. The requested revision will only clarify the conditions when
block valve surveillance testing is required. The performance of
this testing is intended to improve block valve availability and
thereby assure the capability of certain accident mitigation
strategies identified within Abnormal and Emergency Operating
Procedures. Therefore, the requested revision will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The requested revision to our previous TS change request does
not involve a significant reduction in the margin of safety. The
requested revision is intended to help assure block valve
availability to support certain accident mitigation strategies. This
additional assurance of block valve availability and functioning
increases the margin of safety. Accordingly, the requested revision
will not involve a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
NRC Project Director: William H. Bateman.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: December 14, 1994.
Description of amendment request: The proposed amendments would
revise technical specifications related to allowed outage times (AOT)
and surveillance test intervals (STI) for certain actuation
instrumentation in the reactor protection system (RPS), primary
containment isolation system (PCIS), emergency core cooling system
(ECCS), recirculation pump trip, reactor core isolation cooling (RCIC),
control rod withdrawal block, monitoring, and feedwater/main turbine
trip systems. These changes are generally consistent with General
Electric topical reports which have been reviewed and approved by the
NRC. The changes also include revising the Feedwater/Main Turbine Trip
LCO 3.3.8 action statement to achieve consistency with existing
instrumentation LCOs; deleting the surveillance of the APRM Neutron
Flux--High, Setdown functional unit in Operational Condition 1;
revising the applicability of the provisions of Specification 4.0.4 to
several Reactor Protection System and Control Rod Withdrawal Block
Instrumentation surveillance requirements; adding the requirement to
perform shiftly channel checks for applicable RPS, PCIS, ECCS, and RCIC
instrumentation channels equipped with master trip units; and other
changes to correct typographical errors and to delete cycle specific
footnotes which are no longer applicable.
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:

It has been determined that the changes do not constitute a
Significant Hazards Consideration. Based on the criteria for
defining a significant hazards consideration established in 10 CFR
50.92, operation of LaSalle County Station Units 1 and 2 in
accordance with the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
a. The proposed changes increase the STI and AOT for actuation
instrumentation supporting RPS, ECCS, Isolation, CRBF, RCIC, ATWS-
RPT, EOC-RPT, Monitoring, and Feedwater/Main Turbine Trip System
Actuation functions. There are no changes in instrumentation
configuration and function, and no instrumentation setpoints are
changed. Because of this there is no change in the probability of
occurrence of an accident or the consequences of an accident or the
consequences of malfunction of equipment. With respect to the
probability of equipment malfunction, topical reports prepared by GE
demonstrate that there is a reduction in scram frequency for the
RPS, but in the case of the ECCS there is a small increase in the
unavailability of the water injection function. This increase in
unavailability was judged acceptable by GE. The NRC concurred with
this conclusion in its review and approval of the topical reports.
The proposed changes are consistent with the Safety Evaluation
Reports issued for the topical reports.
b. The changes proposed for the Feedwater/Main Turbine Trip LCO
action statements provide actions which are consistent with
presently existing instrumentation LCOs. The design and function of
the feedwater/main turbine trip instrumentation to trip the
feedwater pumps and the main turbine upon detection of a Level 8
event is not altered. The probability and/or consequences of this
moderate frequency transient are not increased.
c. The APRM Neutron Flux--High, Setdown scram setting provides
adequate thermal margin between the setpoint and the safety limits
for operation at low pressure and low flow during a plant startup.
This function remains in effect until the mode switch is placed in
the Run (Operational Condition 1) position, at which time it is
bypassed. Deleting the requirement for the surveillance of the APRM
Neutron Flux--High, Setdown functional unit in Operational Condition
1 is appropriate since its function is not applicable in this mode.
This deletion serves to achieve consistency between Technical
Specification Tables and the Bases section.
d. The changes associated with Specification 4.0.4 are
administrative in nature and are intended to provide the plant
operators with better guidance for its application. In cases where
complete surveillances cannot be achieved, such as during a plant
shutdown, then the required surveillances will be performed within
24 hours of entering the Mode or condition in which the surveillance
is required. The stabilization of the plant will be of primary
consideration. This change does not affect the evaluation for any
accident presented in Chapter 15 of the UFSAR. The APRM Fixed
Neutron Flux--High quarterly functional tests most of the APRM
channel equipment associated with the APRM Neutron Flux--High,
Setdown scram.
Additionally, the expected result of the functional tests
associated with the SRMs, IRMs, and APRMs is to demonstrate the
operability of the instrumentation. Therefore, 24 hours is a
reasonable time to permit the surveillances to be performed upon
entering the mode or condition in which the surveillance is
required.
e. The proposal to include the performance of channel checks as
requirements of technical specifications is administrative in
nature. Presently, channel checks performed for the applicable
analog instrumentation in reactor vessel water level applications is
controlled solely by procedure. Adding this
[[Page 11129]] requirement to the technical specifications provides
for the appropriate controls of the surveillances, above and beyond
that presently controlled by procedure.
f. The proposed administrative changes are offered to correct
typographical errors and delete cycle specific footnotes which are
no longer applicable. The nature of the changes precludes them from
impacting previously analyzed accidents.
The proposed changes therefore 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 because:
a. The proposed changes increase the STI and AOT for certain
actuation instrumentation in the RPS, ECCS, Isolation, CRBF, RCIC,
ATWS-RPT, EOC-RPT, Monitoring, and Feedwater/Main Turbine Trip
systems. There are no changes in instrumentation configuration and
function, and no instrumentation setpoints are changed.
b. The changes to the Feedwater/Main Turbine Trip LCO action
statements allow the plant operators a maximum degree of operational
flexibility, while maintaining the instrumentation and protection
needed for terminating the feedwater controller failure transient.
The single failure proof criterion of the level sensors is
maintained, and the logic of the protective instrumentation is not
compromised. The changes to the LCO action statements do not
constitute a change to the facility or its operation as described in
the Safety Analysis Report.
c. Deleting the requirement for surveilling the APRM Neutron
Flux--High, Setdown functional unit in Operating Condition 1 does
not degrade thermal margins. The margin accommodates the anticipated
maneuvers associated with plant power ascension. During a plant
shutdown, rod insertion maneuvers, recirculation flow reduction, and
xenon build-in all contribute to negative reactivity insertion which
precludes the degradation and violation of thermal margins. The
functions of the APRMs required to be OPERABLE in Operational
Condition 1 which are in effect remain to ensure that reactor core
thermal margins are not compromised.
d. The conduct of neutron instrument functional tests in the
plant mode or condition in which the trips are applicable eliminates
unnecessary testing during normal plant operations. The expected
result of the functional testing is to demonstrate the operability
of the instruments. The failure of any single instrument channel
will neither cause nor prevent either a reactor scram or a control
rod block.
e. Including the performance of channel checks for the
applicable analog instrumentation as part of the technical
specifications transfers control of the required surveillances from
procedure to the technical specifications, as appropriate. The
administrative nature of this change does not alter the functions,
setpoints, or configuration of the associated instrumentation.
f. The administrative nature of the changes prevents them from
affecting the functions, setpoints, or configuration of the
associated instrumentation from being affected by the changes.
The proposed changes do not create the possibility for an
accident or malfunction of a different type than any previously
evaluated in the UFSAR.
(3) Involve a significant reduction in the margin of safety
because:
a. Setpoints are based upon the drift occurring during an 18
month calibration interval. The bases in the Technical
Specifications either do not discuss STI, or state ``* * * one
channel may be inoperable for brief intervals to conduct required
surveillance.'' The proposed changes are bounded by the analyses of
the topical reports. These analyses, which were prepared by GE and
approved by the NRC, examined the effects of extending STI and AOT
and found that the proposed changes would not involve a significant
reduction in the margin of safety.
b. The proposed changes to the turbine trip LCO action
statements do not change any of the settings of the Level 8
setpoints. The single failure criteria of the multiple level sensors
which sense and detect the Level 8 setpoint remains intact. The LCO
maintains the requirement that no single instrument failure will
prevent the feedwater pump turbines and main turbine trip on a valid
Level 8 signal. Scram trip signals from the turbine retain the
design feature that a single failure will neither initiate nor
impede the initiation of a reactor scram (trip).
c. The setting, function, and conditional requirements of the
APRM Neutron Flux--High, Setdown function are not altered. This
change serves to achieve consistency between two Technical
Specifications Tables. This eliminates the need for surveilling a
function in a mode which is not applicable. The functions of the
APRMs required to be OPERABLE in Operational Condition 1 remain to
ensure that reactor core thermal margins are not compromised.
d. The reference to 4.0.4 applicability will assist to ensure
consistent interpretation of the technical specifications by the
plant operators. This assists in ensuring that the plant is operated
within technical specification limitations. This change does not
affect trip instrumentation setpoints, and the scram function of the
RPS is assured by the weekly functional testing of the Manual Scram.
e. Including the instrumentation channel checks as part of
technical specification requirements provides an appropriately
regimented method of controlling the conduct of the surveillances.
None of the functions, setpoints, or configuration of the associated
analog instrumentation is affected by this administrative change.
f. The administrative nature of the changes serves to provide
more concise guidance to the plant operating staff, and as such do
not impact the safety margin.
The proposed changes do not significantly reduce the margin of
safety as defined in the basis for any Technical Specification.

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: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: January 13, 1995.
Description of amendment request: The proposed amendments would
revise the pressure alarm setpoint allowable values for the emergency
core cooling system (ECCS) and reactor core isolation cooling (RCIC)
system ``keep filled'' pressure instrumentation channels. The purpose
of the proposed change is to lower the setpoint allowable values for
these parameters to more realistic values based upon calculations
performed by the licensee reflecting design changes and system
performance. Also, the term ``setpoint'' is being changed to ``setpoint
allowable value'' to clarify the use of the values. Additionally, two
administrative/editorial changes are included to delete technical
specification footnotes which are no longer applicable.
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:

Commonwealth Edison has evaluated the proposed Technical
Specification Amendment and determined that it does not represent a
significant hazards consideration. Based on the criteria for
defining a significant hazards consideration established in 10 CFR
50.92, operation of LaSalle County Station Units 1 and 2 in
accordance with the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
a. The proposed change in the technical specification allowable
values for the ECCS and RCIC discharge line ``keep filled'' alarm
instrument channels does not change the design bases or function of
these systems as described in the technical specifications and
UFSAR. An analysis performed by engineering demonstrates that the
proposed allowable values are sufficient for verifying that the ECCS
and RCIC pump discharge lines are full of water. In addition,
setpoint [[Page 11130]] calculations have been performed to verify
that sufficient margin exists between the recommended calibration
setpoints and the analytical limits for these instrument channels to
account for all applicable instrument errors. This provides high
assurance that the trip setpoints of these instrument channels will
not drop below the minimum required value. The ``keep filled''
instrumentation is not a factor in the assumptions of any accidents,
thus, the probability of analyzed accidents is not increased.
b. The proposed technical specification amendment does not
revise the configuration of the ECCS and RCIC discharge line ``keep
filled'' instrument channels or sensing lines. The proposed setpoint
allowable values and associated calibration setpoints are within the
calibration ranges of the existing pressure switches. Thus,
implementation of the proposed amendment does not involve any
physical alterations to the plant except for the recalibration of
the pressure switches to the new calibration setpoints.
c. The ECCS and RCIC discharge line ``keep filled'' instrument
channels only perform a monitoring function. Other than ensuring
system readiness they do not perform a function important to safety.
Thus, the probability of a ECCS or RCIC failure is not increased
since the operation and function of the ECCS and RCIC discharge line
fill systems is not affected by this change.
d. The failure of a ECCS or RCIC discharge line fill system will
not go undetected by the proposed change, since water leg pump trips
are annunciated in the control room. In addition, quarterly
surveillances are performed on these pumps to check for degradation.
e. The ECCS and RCIC discharge line fill systems are not used to
mitigate the consequences of an accident or transient. These systems
are not required after the ECCS and RCIC pumps are activated.
Therefore, the proposed change does not cause an 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 because: This
technical specification amendment only lowers the trip setpoint
allowable values for the ECCS and RCIC discharge line ``keep
filled'' alarm instrumentation channels. As described above, the
proposed setpoint allowable values are sufficient for verifying that
the ECCS and RCIC discharge lines are full of water. Thus, the
probability of a water hammer occurring during system activation for
a surveillance test is not increased. In addition, each instrument
channel is independent from the other channels so that a failure in
one channel will not propagate to another channel. Therefore, the
operation of the facility in accordance with the proposed amendment
does not create the possibility of a new or different kind of
accident.
(3) Involve a significant reduction in the margin of safety
because: The margin of safety is not affected by this amendment,
because this change involves monitoring instrumentation only. The
purpose of the ECCS and RCIC discharge line ``keep filled'' alarms
is to alert the operators when a ECCS or RCIC system may not be
operable due to empty or partially empty discharge lines. The
proposed amendment does not alter or degrade this function, since
the new setpoint allowable values are adequate for verifying that
the discharge lines are full of water. Therefore the operation of
the facility in accordance with the proposed amendment does not
involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
NRC Project Director: Robert A. Capra.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan

Date of amendment request: January 13, 1995
Description of amendment request: The proposed amendment would
modify the required settings, and allowable ``as found'' and ``as
left'' tolerances for the primary and secondary safety valves. The
proposed limits would allow installed primary and secondary valve
settings to be within a 3% tolerance of their nominal settings, but
would require returning the valve settings to within 1% of the nominal
settings if the valves are removed from the piping for maintenance or
testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The following evaluation supports the finding that operation of
the facility in accordance with the proposed technical specification
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the Technical Specifications increases
the acceptable as found tolerance for the pressurizer safety valves.
The most limiting overpressure event, loss of external load, has
been analyzed to account for this change. The loss of external load
analysis was performed using a conservative 25% steam generator tube
plugging and an initial pressurizer level of 67.8% (providing an
approximate 10% conservative margin above programmed pressurizer
level for full power). Primary and secondary safety valve
accumulation was conservatively accounted for and the setpoint
tolerance of +3% was assumed. Reactor trip on turbine trip was
assumed to be disabled and the atmospheric dump valves were assumed
unavailable. The results of the analysis demonstrated primary and
secondary system pressures within 110% of design pressures.
Therefore, the consequences of overpressurization events will not be
significantly increased with a +3% tolerance on the primary safety
valve setpoints. The proposed Technical Specifications change will
not affect normal plant operation and will not increase the
probability of an accident.
A review of all DNB [departure from nucleate boiling] analyses
was performed to ensure that predicted pressurizer pressures for
those analyses would not be affected by a -3% tolerance on the
lowest setpoint valve. The DNB analyses for which significant
primary system pressure increases were predicted do not result in
pressures high enough to lift the pressurizer safety valves with the
proposed tolerance. A conservative DNB analysis that bounds the
consequences of inadvertent opening of a pressurizer safety valve
has also been previously performed with predicted acceptable
results. If a pressurizer safety valve were to stick open, the
consequences would be bounded by the small break LOCA [loss-of-
coolant accident] analysis. Therefore, the consequences due to a -3%
tolerance on the primary safety valve setpoints will not increase
the consequences or probability of an accident.
The proposed revision removes the requirement for one operable
pressurizer safety valve to be installed whenever the reactor head
is on the vessel. Instead, proposed Specification 3.1.7.1 requires
all pressurizer safety valves to be operable above cold shutdown,
and overpressure protection during cold shutdown is provided by
existing Specification 3.1.8.2, Power Operated Relief Valves.
The proposed Technical Specifications change also lists the lift
settings for each of the primary and secondary system safety valves.
This change will not affect the operation or function of the valves.
Therefore, the probability and consequences of previously evaluated
accidents will not be increased.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed changes to Technical Specifications will not affect
the manner in which the plant operates. The proposed increase in
pressurizer safety valve lift setting tolerance could change the
pressure at which the valves open in an overpressurization event,
but would not create the possibility of a new or different kind of
accident. Since Technical Specification 3.1.8 addresses primary
system overpressurization during cold shutdown, the proposed removal
of the requirement for an operable pressurizer safety valve to be
installed whenever the reactor head is on the vessel will not create
[[Page 11131]] the possibility of a new overpressurization event
during cold shutdown. The proposed change to list the lift settings
for the individual primary and secondary safety valves will have no
effect on the safety function of the valves. Therefore, the proposed
changes will 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 proposed changes to Technical Specifications do not affect
the DNB analyses that have been previously performed. The most
limiting overpressurization event, loss of external load, has been
conservatively analyzed accounting for the proposed changes and
demonstrated that the primary and secondary system pressures remain
within 110% of the design pressures. Overpressurization during cold
shutdown is addressed by Technical Specification 3.1.8. 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: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: John N. Hannon.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan.

Date of amendment request: February 10, 1995.
Description of amendment request: The proposed amendment would
modify the Technical Specifications to allow a one time deferral of
several 18-month interval surveillance tests until the upcoming
scheduled refueling outage to avoid the necessity of imposing a plant
shutdown solely for the sake of their performance.
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 following evaluation supports the finding that operation of
the facility in accordance with the proposed Technical
Specifications (TS) would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Deferring surveillance testing will introduce no new operating
conditions, change no equipment operating procedures, and change no
plant systems or equipment. Therefore, operation of the facility in
accordance with the proposed TS would not result in a significant
increase in the probability of an accident previously evaluated.
Deferring surveillance testing of snubbers and instrument
channels could allow minor degradations of snubber condition or
small changes in instrument setpoints or calibration to progress
some amount beyond that point which would occur with a shorter
surveillance interval. A review of the recent test history for the
subject surveillance indicates that no significant snubber
degradation or instrument drift was found. It is not expected that,
even with the proposed surveillance deferral, snubber conditions or
instrument settings will be found to exceed conditions allowable by
the Technical Specifications. Therefore, operation of the facility
in accordance with the proposed TS would not result in a significant
increase in the consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
Deferring surveillance testing will introduce no new operating
conditions, change no equipment operating procedures, and change no
plant systems or equipment. Therefore, operation of the facility in
accordance with the proposed TS would not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
A review of past performance of the subject surveillance tests
indicate that the requested deferral of testing would not have a
significant effect on the results of the tests when they are
performed prior to the startup for cycle 12. Most of the affected
instrumentation is monitored each shift by channel checks, which
would disclose major failures or significant drift. Therefore,
operation of the facility in accordance with the proposed TS would
not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: John N. Hannon.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: November 2, 1994.
Description of amendment request: The proposed amendment would
delete the content of the Appendix B, Environmental Protection Plan
(EPP) and modify License Conditions 2.C.(2) to delete that portion
which refers to the EPP. Specifically, the requirements for non-
radiological environmental monitoring have been completed. The
radiological environmental monitoring requirements have been
incorporated into Appendix A (the Technical Specifications). There
would be no impact on the continued safety of the McGuire station by
deleting Appendix B.
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:

Deletion of the Environmental Protection Plan and modifying
License Condition 2.C.(2) will have no impact on the probability or
consequences of an accident previously evaluated because the changes
will not have any impact upon the design or operation of any plant
systems or components.
The proposed revision will not create the possibility of a new
or different kind of accident from any previously evaluated because
the revision is administrative in nature and will not change the
types and amounts of effluent that will be released.
The proposed revision will not reduce a margin of safety because
it is administrative in nature and will not effect the margin of
safety as defined in the basis for any Technical Specifications.
Accordingly, this proposed changes does not involve a
significant hazard.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow. [[Page 11132]]

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: January 18, 1995.
Description of amendment request: The proposed amendments would
relocate the requirements for the seismic instrumentation,
meteorological instrumentation, and loose-part detection system from
the Technical Specifications to the Selected Licensee Commitment (SCL)
Manual. This will allow future changes to these controls to be
performed under the provisions of 10 CFR 50.59. No changes are being
made to the technical content of the affected Technical Specification
pages.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Relocation of the affected TS sections to the SLC Manual
will have no effect on the probability of any accident occurring. In
addition, the consequences of an accident will not be impacted since
the above instrumentation will continue to be utilized in the same
manner as before. No impact on the plant response to accidents will
be created.

Criterion 2

The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No new accident causal mechanisms will be created as a
result of relocating the affected TS requirements to the SLC Manual.
Plant operation will not be affected by the proposed amendments and
no new failure modes will be created.

Criterion 3

The requested amendments will not involve a significant
reduction in a margin of safety. No impact upon any plant safety
margins will be created. Relocation of the affected TS requirements
to the SLC Manual is consistent with the content of the Westinghouse
RSTS [Revised Standard Technical Specifications], as the NRC did not
require technical specification controls for the affected
instrumentation in the RSTS. The proposed amendments are consistent
with the NRC philosophy of encouraging utilities to propose
amendments that are consistent with the content of the RSTS.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: January 18, 1995.
Description of amendment request: The amendments would revise
Technical Specification Table 4.3-3 to allow the analog channel
operational test interval for radiation monitoring instrumentation to
be increased from monthly to quarterly. The proposed amendment changes
would be consistent with the guidance in Generic Letter 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Decreasing the frequency of the radiation monitor analog
channel operational test from monthly to quarterly will have no
impact upon the probability or any accident, since the radiation
monitors are not accident initiating equipment. Analysis of the
previous test data * * * shows that no significant degradation of
performance is to be expected by the decrease in frequency.
Therefore, the requested amendments will have no adverse impact upon
the consequences of any accident.

Criterion 2

The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, the radiation monitors are not accident
initiating equipment. No new failure modes can be created from an
accident standpoint. The plant will not be operated in a different
manner.

Criterion 3

The requested amendments will not involve a significant
reduction in a margin of safety. Plant safety margins will be
unaffected by the proposed changes. No safety equipment which is
taken credit for in accident analyses will be affected by the
requested amendments. The availability of the affected radiation
monitors will be increased as a result of the proposed amendments
because the monitors will not have to be made unavailable for
testing as frequently. In addition, radiation monitor operating
experience supports the proposed amendments. Finally, the proposed
amendments are consistent with the NRC position and guidance set
forth in NUREG-1366 and Generic Letter 93-05.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: January 20, 1995.
Description of amendment request: The proposed amendments will
relocate the operability requirements for the INCORE DETECTORS (TS 3/
4.3.3.2) to the Updated Final Safety Analysis Report, and revise Linear
Heat Rate surveillance 4.2.1.4, and Special Test Exceptions
surveillances 4.10.2.2, 4.10.4.2 (Unit 2 only), and 4.10.5.2,
accordingly.
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 changes are administrative in nature in that the
specifications for [[Page 11133]] operation and surveillance of the
Incore Instrumentation (ICI) System will be relocated from the
Technical Specifications to the Updated Final Safety Analysis Report
for St. Lucie Unit 1 and Unit 2. Changes to the system will be
controlled by 10 CFR 50.59, and the safety analysis report is
required to be updated pursuant to 10 CFR 50.71(e). Relocation of
these requirements to the UFSAR is consistent with the NRC ``Final
Policy Statement on Technical Specifications Improvements for
Nuclear Power Reactors'' published in the Federal Register (58 FR
39132) dated July 22, 1993.
Incore instrumentation is not an accident initiator nor a part
of the success path(s) which function to mitigate accidents
evaluated in the plant safety analyses. The proposed technical
specification change does not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor do the changes
alter any assumptions or conditions in any of the plant accident
analyses. Therefore, 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.
(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 amendment to relocate the existing Technical
Specification requirements for the Incore Instrumentation System to
the Updated Final Safety Analysis Report will not change the
physical plant or the modes of plant operation defined in the
Facility License. The change does not involve the addition or
modification of equipment nor does it alter the design or operation
of plant systems. Therefore, 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.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed changes are administrative in nature in that
operating and surveillance requirements for the Incore
Instrumentation System will be relocated from the Technical
Specifications to the Updated Final Safety Analysis Report for St.
Lucie Unit 1 and Unit 2. The ICI system is not used to actuate
safety-related equipment, provide interlocks, or otherwise perform
automatic plant control functions. The system is used to monitor
core power distribution parameters whose limits do involve a margin
of safety; however, the ICI system itself makes no contribution to
that margin of safety, and the power distribution limits will not be
changed by the proposed amendment. Therefore, operation of the
facility in accordance with the proposed amendment would not involve
a significant reduction in a margin of safety.
Based on the above discussion and the supporting Evaluation of
Technical Specification changes, FPL has determined that the
proposed license amendment involves no significant hazards
consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L 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: January 17, 1995.
Description of amendment request: The licensee proposes to revise
the technical specifications to reference Topical Report NF-TR-95-01 as
the documentation of the licensee's proficiency in performing certain
reload design calculations once the NRC has evaluated and approved NR-
TR-95-01.
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 addition of the reference to FPL [Florida Power and Light
Company] topical report which demonstrates FPL's ability to perform
certain reload design calculations for Turkey Point Units 3 and 4 is
administrative in nature and has no impact on the probability or
consequences of any Design Bases Event (DBE) occurrences previously
evaluated. The reload design calculations will be performed using
methodologies and computer codes approved by the NRC and poses no
increase in the probability or consequences of any accident
previously evaluated.
The Core Operating Limits Report (COLR) parameters will be
evaluated every cycle to ensure proper compliance with the Updated
Final Safety Analysis Report (UFSAR). These limits will be evaluated
in accordance with 10 CFR [Section] 50.59, which ensures that the
reload will not involve an increase in the probability of
occurrences or consequences of an accident previously evaluated.
Title 10 CFR [Section] 50.59 (2) states that a proposed change
involves an unreviewed safety question (i) if the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety previously evaluated in the safety
analysis report may be increased. Consequently, since any change to
the reload core design analysis must be evaluated relative to the
more restrictive evaluation criterion of 10 CFR [Section] 50.59,
then 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
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The addition of the reference to FPL topical report which
demonstrates FPL's ability to perform certain reload design
calculations for Turkey Point Units 3 and 4 is administrative in
nature and has no impact, nor does it contribute in any way to 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 events are introduced as a result of the
proposed change.
The generation of the Axial Flux Difference, Rod Bank Insertion
limits and K(Z) curve will be performed using NRC-approved
methodology and are submitted to the NRC, as a revision to the COLR,
to allow the NRC staff to trend. The Technical Specifications will
continue to require operation within the core operating limits and
appropriate actions will be taken if these limits are exceeded.
Title 10 CFR [Section] 50.59 permits a licensee to make changes
in the facility as described in the safety analysis report without
prior Commission approval, provided that the proposed changes does
not involve an unreviewed safety question. 10 CFR [Section] 50.59
(2) states that a proposed change involves an unreviewed safety
question (ii) if a possibility for an accident or malfunction of a
different type than any evaluated previously in the safety analysis
report may be created. Consequently, since any change to the reload
core design analysis must be evaluated relative to the more
restrictive evaluation criterion of 10 CFR [Section] 50.59, then
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.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The margin of safety is not affected by FPL performing the
reload design calculations for Turkey Point Units 3 and 4. The
supporting Technical Specification values are defined by the
accident analyses which are performed to conservatively bound the
operating conditions defined by the Technical Specifications. The
development of the limits for future reloads will continue to
conform to the methodology described in NRC approved documentation.
In addition, each future reload will involve a 10 CFR [Section]
50.59 [[Page 11134]] review to assure that operation of the units
within the cycle specific limits will not involve a reduction in a
margin of safety. 10 CFR [Section] 50.59 (2) states that a proposed
change involves an unreviewed safety question (iii) if the margin of
safety as defined in the basis for any technical specification is
reduced. Consequently, since any change to the reload core design
analysis must be evaluated relative to the more restrictive
evaluation criterion of 10 CFR [Section] 50.59, then operation of
the facility in accordance with the proposed amendments would not
involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration. The
NRC staff, however, considers that the licensee's statements relative
to 10 CFR Section 50.59 evaluations to be performed in the future are
not relevant to the proposed no significant hazards determination.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: David B. Matthews.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa

Date of amendment request: October 28, 1994.
Description of amendment request: The proposed amendment revises
the Duane Arnold Energy Center (DAEC) Operating License by deleting a
condition of the license that requires a ``Plan for Integrating
Scheduling of Plant Modifications for the Duane Arnold Energy Center''
(the Plan).
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 provided below:

(1) The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. No physical changes to the facility will occur
as a result of this amendment. Work activities will continue to
receive the appropriate level of review in accordance with DAEC
procedures and practices. The organizational structure that controls
and manages these activities remains unchanged and will assure that
activities are prioritized and performed in a manner consistent with
plant safety. The proposed amendment removes an administrative
burden that is no longer required.
(2) The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No changes to the physical design and/or operation of the
plant will occur as a result of this amendment. The processes by
which activities are planned, prioritized, and controlled are not
affected. The appropriate level of technical review and management
oversight continue to be performed in accordance with existing
procedures and practices to assure that activities are performed in
a manner consistent with plant safety.
(3) The proposed amendment does not involve a significant
reduction in a margin of safety. As stated earlier, no changes to
the physical design and/or operation of any plant systems will occur
as a result of this amendment. Work activities will continue to
receive the appropriate technical review and management oversight to
assure that activities are prioritized and performed in a manner
consistent with plant safety. The amendment removes an
administrative burden that is no longer required.
Based on the above, we have determined that the proposed
amendment will not involve a significant hazards consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis
& Bouckins, 1800 M Street NW., Washington, DC 20036.
NRC Project Director: Leif J. Norrholm.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: January 24, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.4.1, ``Leakage Rate,'' to reduce the
allowable leakage rate of the reactor building from 2000 cubic feet per
minute (cfm) to 1600 cfm.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Secondary containment and RBEVS [Reactor Building Emergency
Ventilation System] are not initiators or precursors to an accident.
Secondary containment provides a pressure boundary, with limited in-
leakage, for the purpose of preventing a ground level unfiltered
release of radioactivity. RBEVS responds to accidents involving
release of radioactivity to the secondary containment by maintaining
a negative pressure inside secondary containment and by providing an
elevated release. Therefore, a change to the Reactor Building
leakage rate cannot affect the probability of an accident previously
evaluated.
Although the proposed change reduces the Reactor Building
leakage rate from 2000 cfm to 1600 cfm consistent with system
design, there is no effect on the radiological consequences of any
previously analyzed accident since the radiological analysis does
not assume exfiltration. Therefore, the Technical Specification
change does not significantly increase the consequences of a
previously evaluated accident.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change to the Reactor Building leakage rate from
2000 cfm to 1600 cfm does not involve any accident precursors or
initiators. During an accident involving a release of radioactivity
to the secondary containment, the RBEVS would be operable and
provide filtration of containment atmosphere prior to release to the
environment. This change does not involve any physical modifications
to the system, thus the system will operate as designed. Therefore,
the proposed Technical Specification change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed change in Reactor Building in-leakage from 2000 cfm
to 1600 cfm in Specification 3.4.1 and the associated basis is to be
consistent with system design and reflect the leakage rate
associated with approximately one building air volume change per
day. The resulting accident analysis remains unchanged since the
radiological analysis does not assume any exfiltration. Therefore,
the proposed change will not involve a significant reduction in the
margin of safety as defined in the basis for any Technical
Specification.
Therefore, as determined by the above analysis, this proposed
amendment involves no significant hazards consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request [[Page 11135]] involves no significant hazards
consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Ledyard B. Marsh.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: February 1, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.6.13, ``Remote Shutdown Panels.''
TS 3.6.13 currently requires that if the valve controls or monitoring
instrumentation on the Remote Shutdown Panels are inoperable, they must
be restored to an operable status within 24 hours or the plant shall be
shut down. The proposed change would require inoperable valve control
functions be restored to an operable status within 30 days or the plant
shall be shut down. The proposed change would also specify that
required inoperable monitoring instrumentation functions be restored to
an operable status within 30 days or that an alternate method of
monitoring the parameter be established within 30 days and the required
function be restored to an operable status within 90 days or the plant
shall be shut down.
The proposed amendment would also make minor editorial changes to
TS Table 3.6.13-1 so that the table entries would be consistent with
the proposed revisions to TS 3.6.13.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The remote shutdown panel monitoring instruments and controls
are not initiators or precursors to an accident. The remote shutdown
panels provide the operator with sufficient monitoring instruments
and controls to place and maintain the plant in a safe shutdown
condition from a location other than the control room. Therefore,
the proposed changes to Specification 3.6.13, ``Remote Shutdown
Panels,'' cannot affect the probability of a previously evaluated
accident.
The proposed changes, in part, require that one channel (on
either panel) for each function be operable. This change could
potentially avoid an unnecessary plant shutdown without affecting an
operator's ability to cope with a control room evacuation. One
channel of each function is adequate to assure a safe shutdown. The
proposed changes would also allow 30 days to restore an inoperable
function to an operable status. As indicated in the ITS [Improved
Standard Technical Specifications], the allowed time of 30 days is
acceptable based on operating experience and the low probability of
an event that would require evacuation of the control room. With one
or more monitoring instrument functions inoperable, the proposed
change gives an operator an additional option. Specifically, the
operator is allowed 30 days to establish an alternate method of
monitoring the parameter and 90 days to restore the function to
operable status. The use of an alternate method is acceptable since
it will provide the operator with indication of the parameter of
interest. The remote shutdown panels will not be required to be
operable in hot shutdown because the plant is already subcritical
and in a condition of reduced reactor coolant inventory energy.
Because this Specification no longer applies to hot shutdown and to
be consistent with the guidance provided in the ITS, Specification
3.6.13.d will require that the plant be brought to a hot shutdown
condition (versus cold shutdown condition) in 12 hours. As indicated
in the ITS, the 12-hour completion time is reasonable based on
operating experience. The Bases Section to 3.6.13 and 4.6.13 was
revised to be consistent with the proposed changes to the
Specification. The Bases currently indicates that one remote
shutdown panel is required to be operable. As explained above, one
channel of each required function is required to maintain remote
shutdown operability. In summary, the proposed changes will not
affect the ability of the Remote Shutdown System to provide the
operator with sufficient instrumentation and controls to place and
maintain the plant in a safe shutdown condition from a location
other than the control room. Therefore, the consequences of an event
requiring a control room evacuation will not significantly increase.
Editorial changes were made to Table 3.6.13-1 to be consistent
with the changes made to the Specification. Specifically, the word
``INSTRUMENT'' was changed to ``FUNCTION'' and the words ``PANEL
MONITORING'' were changed to the words ``PANELS FUNCTIONS.'' These
changes make it clear that one channel of each function, on either
panel is acceptable to maintain operability. The emergency condenser
condensate return valve control and motor-operated steam supply
valves control were relocated from Specification 3.6.13.b to Table
3.6.13-1 to be consistent with the proposed changes.
Based on the above, the consequences of an accident previously
evaluated are not significantly increased.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The changes do not introduce any new accident precursors and do
not involve any alterations to plant configurations which could
initiate a new or different kind of accident. The proposed changes
require that one channel of each function be operable to assure the
remote shutdown panels can meet their intended function. No changes
have been made which will affect the operation of the remote
shutdown panels in a way which would create a new or different kind
of accident. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any
previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed changes will not affect the ability of the Remote
Shutdown System to provide the operator with sufficient
instrumentation and controls to place and maintain the plant in a
safe shutdown condition from a location other than the control room.
The ability to respond to a control room evacuation is maintained
with one channel operable for each required function. The allowed
outage time of 30 days is acceptable based on operating experience
and the low probability of an event requiring control room
evacuation. 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, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Ledyard B. Marsh.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: January 10, 1995.
Description of amendment request: The proposed amendment request
would revise Technical Specifications by deleting the power range,
neutron flux, high negative rate trip from Tables 2.2-1, 3.3-1, and
4.3-1, and delete the associated Bases Section 2.0. [[Page 11136]]
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:

* * * The proposed changes would not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The deletion of the power range, neutron flux, high negative
rate trip will not adversely affect plant operations. As has been
presented and accepted by the NRC Staff in previous docketed
correspondence, the dropped RCCA [rod cluster control assembly]
accident analysis does not rely on this trip to safely shut down the
plant. The safety analysis of the plant is unaffected by the
proposed changes. Since the safety analysis is unaffected, the
calculated radiologicalreleases associated with the analysis are not
affected. Therefore, the proposed changes will not increase the
probability or consequences of an accident previously evaluated.
The reactor trip system is used to mitigate accidents. There
have been instances, during calibration of these units, where a
single channel has generated a trip signal. Leaving this in place
when it is not necessary could, therefore, cause a reactor trip. The
deletion of one trip function will, therefore, slightly decease, not
increase, this probability.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The reactor trip system is used to mitigate accidents, and the
only way that it can initiate an event is by causing the reactor to
trip when it is unnecessary. This possibility of the generation of a
false trip signal has already been evaluated in the safety analysis.
This modification will physically remove or disable the power range,
neutron flux trip and will therefore decrease the possibility for
the generation of a false trip signal. Therefore, the proposed
change cannot create a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change which deletes the power range, neutron flux,
high negative rate trip will have no impact on the margin of safety.
The current safety analysis for Millstone Unit No. 3 does not credit
this trip for any events; therefore, removal of this trip from the
technical specifications will not affect the margin of safety for
any analyzed events.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: January 23, 1995.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) by 1) adding a new Section 3/
4.5.5 which provides a limiting condition for operation, an action
statement, a surveillance requirement, and a corresponding bases
section, for the trisodium phosphate (TSP) baskets which will be
installed in the next refueling outage; 2) deleting Section 3/4.6.2.3
and Bases 3/4.6.2.3 related to the spray additive system which are no
longer needed since the chemical addition tank is being abandoned; and
3) updating Index Pages viii, ix, and xiv to reflect the above changes.
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:

* * * The proposed changes do not involve an SHC because the
changes would not:
1. Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
The plant change affects the chemical composition of the QSS
[quench spray system] flow and the method of sump pH control, which
are important for containment heat removal/pressure mitigation (MSLB
and LOCA) [main steamline break and loss-of coolant accident] and
fission product removal (LOCA). However, this change does not affect
the probability of occurrence of these accidents. Since the TSP
baskets are passive devices located inside the containment, they
cannot initiate a transient or affect the probability of occurrence
of any previously evaluated accident.
The design change will not adversely affect the radiological
doses for the DBA [design basis accident] LOCA at the Exclusion Area
Boundary, Low Population Zone, Millstone Unit No. 3 Control Room,
Millstone Unit No. 2 Control Room, and the Millstone Technical
Support Center. Also, the change will not adversely affect the
calculated peak clad temperature (PCT) for the DBA LOCA.
2. Create the Possibility of a New or Different Kind of Accident
from any Previously Analyzed.
The change does not create a malfunction that is different from
those previously evaluated. The TSP baskets are passive devices that
have minimal impact on any other systems except through water
chemistry. The change in water chemistry does not adversely affect
any safety systems. The installation of the TSP baskets and the
abandonment of the CAT [chemical addition tank] will not change the
probability of a malfunction of safety-related equipment.
Potential malfunctions relating to the TSP powder, the 12
baskets which hold the TSP powder, the QSS and other systems, and
equipment credited in the safety analysis were evaluated and
determined not to be adversely affected by the change. Additionally,
the transient pH behavior of the spray flow will not adversely
affect metals, coatings and elastomers in the containment, and the
performance of associated safety functions is not affected.
Finally, the change in the chemical composition of the QSS
solution will not affect the operability of this system or its
ability for containment heat removal and pressure mitigation.
3. Involve a Significant Reduction in the Margin of Safety.
The design changes do not adversely affect the ability of the
QSS to perform the function of containment heat removal, pressure
mitigation and fission product (iodine) retention. The design
changes do not adversely affect any equipment credited in the safety
analysis. Also, the design changes to not increase the calculated
peak clad temperature (PCT) or the offsite doses due to the design
basis LOCA. Therefore, there is no impact on the margin of safety as
specified 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: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: January 24, 1995. [[Page 11137]]
Description of amendment request: The amendment request would
revise the Technical Specification Section 3.2.3.1.a and Table 2.2-1 to
decrease the acceptance criterion for measured reactor coolant system
(RCS) flow rate from 387,480 gallons per minute (gpm) to 371,920 gpm.
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:

* * * The proposed changes do not involve an SHC because the
changes would not:
1. Involve a Significant Increase in the Probability or
Consequence of an Accident Previously Evaluated.
An evaluation of the 4% decrease in the RCS total flow rate
limit has shown that the change does not significantly impact the
design basis analyses. Therefore, the change will not increase the
consequences of an accident previously evaluated.
There are no actual plant changes that will result from this
technical specification change. Instead, the technical specification
requirement for minimum total RCS flow rate is being changed to
provide operational benefit without compromising safety. Since there
are no plant changes, there is no effect on the probability of
occurrence of previously evaluated accidents.
The change will have a negligible impact on the small break loss
of coolant accident (LOCA) and large break LOCA analyses. The PCT
[peak cladding temperature] acceptance criteria will continue to be
met with the assumption of a 4% reduction in RCS flow rate.
For the steam generator tube rupture event, both the FSAR [Final
Safety Analysis Report] offsite dose analysis and the margin of
steam generator (SG) overfill were evaluated. It was determined that
the 4% reduction in RCS flow rate will not adversely affect the
offsite doses or the margin to SG overfill and, therefore, the FSAR
conclusions remain unchanged.
In the evaluation of non-LOCA transients, the DNB [departure
from nucleate boiling] is the most affected parameter due to a
change in flow rate. It was concluded that the 4% reduction in RCS
flow was acceptable and there was margin to the DNB limit.
It is concluded that there is sufficient margin to the system
pressure, PCT and DNB limits to offset the effect of the 4% flow
rate decrease and the calculated radiological releases associated
with the analysis are not affected. Therefore, there is no effect on
the consequences of previously evaluated accidents.
2. Create the Possibility of a New or Different Kind of Accident
from any Previously Analyzed.
The low loop flow trip setpoint specified in Technical
Specification Table 2.2-1 is set as a fraction of total flow. The
flow fraction is not being changed and no hardware changes are
required due to the reduction in minimum flow. Also, the reduction
in minimum flow will not change the operation of any plant equipment
and it does not modify plant operation.
Therefore, the reduction in minimum flow does not introduce any
new failure modes or malfunctions and it does not create the
potential for a new unanalyzed accident.
3. Involve a Significant Reduction in the Margin of Safety.
The proposed 4% decrease in the technical specification limit
for total RCS flow rate will not adversely affect the results of the
FSAR accident analysis, and it is concluded that this change is
safe. The change does not adversely affect any equipment credited in
the safety analysis, and it does not affect the probability of
occurrence of any plant accident. Also, the change has a negligible
impact on the PCT, and it does not increase the offsite doses or
decrease the DNB below its acceptance limit.
Therefore, the change does not have any significant impact on
the protective boundaries, and there is no reduction in the margin
of safety as specified 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: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska

Date of amendment request: January 9, 1995.
Description of amendment request: The proposed amendment to the
technical specifications (TSs) would delete requirements for the toxic
gas monitoring system (TGMS) as contained in TS 2.22 and TS 3.1, Table
3-3, item 29.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The previously evaluated accidents affected by this change are
the on-site and off-site toxic chemical releases. These events have
been re-evaluated for this proposed change and have been shown to
meet the applicable regulatory screening criteria. The deterministic
analyses performed show that the guidelines of Regulatory Guide 1.78
for control room habitability are met for on-site and most off-site
chemicals. On-site chemical sources originally present when the
toxic gas monitoring system was installed have been removed from
site or determined not to exceed the deterministic analysis
screening requirements. For those off-site chemical releases which
did not meet the deterministic screening criteria a probabilistic
analysis was performed. The probabilistic analysis performed in
support of this proposed change shows that the probability of an
off-site chemical release leading to 10 CFR 100 consequences is
orders of magnitude less than the SRP [Standard Review Plan] 2.2.3
guidelines. These results show that there is no significant increase
in the probability or consequences of any accident previously
evaluated.
(2) The proposed changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
Only events involving chemicals for which the TGMS provides an
automatic detection/isolation function are affected by this change.
As stated above, the potential events involving these chemicals have
been re-evaluated using the appropriate regulatory guidance and
shown to satisfy either the deterministic screening criteria of RG
[Regulatory Guide] 1.78, or to be probabilistically insignificant
compared to the guidelines of SRP Section 2.2.3. These results show
that the proposed change will not create the possibility of a new or
different kind of accident from any previously evaluated. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any previously analyzed.
(3) The proposed changes do not involve a significant reduction
in a margin of safety.
The margin of safety is defined by the regulatory basis for the
existing TGMS, namely NUREG-0737, Item III.D.3.4. The analysis
provided to support this proposed change follows the regulatory
guidelines of RG 1.78 and SRP Section 2.2.3, as specified in NUREG-
0737, Item III.D.3.4. The analysis shows that the applicable
regulatory criteria are met and 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: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875
Connecticut [[Page 11138]] Avenue, NW., Washington, DC 20009-5728.
NRC Project Director: Theodore R. Quay.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California

Date of amendment requests: February 6, 1995 (Reference LAR 95-01).
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2, to change TS 3/4.9.14.1, ``Spent Fuel
Assembly Storage,'' TS 3/4.9.14.2, ``Spent Fuel Pool Boron
Concentration,'' TS 5.3.1, ``Reactor Core--Fuel Assemblies,'' and TS
5.6.1, ``Fuel Storage--Criticality,'' and add new TS 3/4.9.14.3,
``Spent Fuel Assembly Storage--Spent Fuel Pool Region 1.'' The specific
TS changes proposed are as follows:
(1) The proposed changes to TS 3/4.9.14 are:
(a) TS 3.9.14.1 and Figure 3.9-2 would be revised to allow the
storage of spent fuel assemblies with initial enrichments up to 5.0
weight percent uranium-235 (U-235) in Region 2 of the spent fuel pool
(SFP). Fuel pellet diameter would be considered in combination with
initial enrichment and cumulative burnup.
(b) Editorial corrections to the titles of TS 3/4.9.14.1 and 3/
4.9.14.2 would be made for consistency with the TS format.
(2) New TS 3/4.9.14.3 would be added. The new TS would include:
(a) Requirements for acceptable fuel storage in Region 1 of the
SFP.
(b) An action statement, similar to that for TS 3.9.14.1, requiring
suspension of all fuel movement and crane operations except to move the
noncomplying fuel assemblies into an acceptable pattern. The action
statement also requires verification of SFP boron concentration at
least once per 8 hours.
(c) A requirement, similar to that for TS 4.9.14.1, for an
evaluation that considers enrichment, boron content, and cumulative
burnup of each fuel assembly before storage in Region 1 of the SFP.
(d) New Figure 3.9-3 for use in determining the acceptability of
storing fuel in Region 1 of the SFP.
(3) The proposed changes to TS 5.3.1 are:
(a) The number of fuel rods in each fuel assembly, nominal length
of each fuel rod, and maximum fuel enrichment would be removed.
(b) The current allowance for fuel rod substitutions as justified
by analysis would be clarified to specify that the analysis be
performed using NRC staff-approved methods.
(c) An allowance to use a limited number of lead test assemblies in
nonlimiting core locations would be added.
(d) The current specification requiring Zircaloy-4 fuel cladding
would be changed to allow Zircaloy-4 or ZIRLO cladding.
(4) The proposed changes to TS 5.6 are:
(a) TS 5.6.1.1 would be renumbered TS 5.6.1 and the word
``borated'' would be replaced with ``unborated.''
(b) A new requirement would be added to specify the maximum fuel
enrichment allowed to be stored in the fuel racks.
(c) TS 5.6.1.2 would be deleted.
(5) The associated Bases would also be appropriately revised.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Analyses were performed to verify that an increase in enrichment
of the fuel from 4.5 weight percent U-235 to 5.0 weight percent U-
235 would not result in an inadvertent criticality event in the new
fuel storage racks or the SFP. The analyses indicate that for the
new fuel racks, the keff will remain below 0.95 if flooded with
non-borated water, and below 0.98 if flooded with optimum-density
aqueous foam. The analyses indicate that for the spent fuel racks,
assuming credit for soluble boron in accident scenarios, the
keff will remain below 0.95 as required.
The increase in the fuel enrichment from 4.5 weight percent U-
235 to 5.0 weight percent U-235 does not change any of the external
dimensional characteristics of the fuel element, the fuel storage
racks, or the SFP itself. The accidents originally evaluated
considered those events that could lead to fuel damage and release
of radioactive material primarily from mechanical means, such as
physical impact on the fuel or the SFP. Because the physical design
and methods of operation are the same as previously evaluated, there
is no change in the probability of occurrence of such events.
The maximum spent fuel gap activity and the resulting offsite
dose consequences after a postulated fuel handling accident are
primarily dependent on fuel burnup, and are not significantly
affected by an increase in fuel enrichment. For up to 5.0 weight
percent U-235 and 60,000 MWD/MTU burnup, NUREG/CR-5009 indicates
that fuel handling accident offsite doses could increase by a factor
of 1.2, which indicates that doses would still remain within 10 CFR
Part 100 limits.
The Generic Letter 90-02 Supplement 1 change to TS 5.3.1
clarifies the requirements associated with fuel reconstitution. It
does not change the methodology that would be used to reconstitute
fuel.
The use of ZIRLO cladding will not increase the probability or
consequences of an accident, since it has improved mechanical
properties such as a lower corrosion rate and reduced radiation-
induced growth.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The physical and mechanical parameters associated with the fuel
assemblies and spent fuel racks are the same as previously
evaluated. Therefore, any malfunctions related to the physical
aspects of fuel storage are the same as previously evaluated.
The conditions for fuel storage in the proposed new TS 3.9.14.3
provide new criteria for locations where a fuel assembly could be
incorrectly placed. However, the incorrect placement of a fuel
assembly has been analyzed, and would not cause an inadvertent
criticality or any other accident.
The change to 5.0 weight percent U-235 does not result in
physical alterations or changes to the operation of the plant, or
change the method by which any safety-related system performs its
function. The use of ZIRLO cladding does not result in a significant
change to the plant.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
The acceptance criteria of a keff of 0.95 (or 0.98 for the
new fuel rack optimum moderation accident) provides the margin to
criticality. Analyses were performed that conclude that the proposed
changes to allow up to 5.0 weight percent U-235 in the new and spent
fuel racks meet the acceptance criteria. The use of ZIRLO cladding
will not reduce the protection of the public health or safety, as
indicated in the NRC's revisions to 10 CFR 50.44 and 50.46 (57 FR
39355).
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
[[Page 11139]] Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: Theodore R. Quay.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California

Date of amendment request: November 23, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications Section VI, ``Waste Disposal
Systems,'' regarding radioactive effluent limitations and the
conditions for automatically pumping the contents of the reactor
caisson sump to the outfall canal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed revisions to the HBPP Technical Specifications
remove the ambiguity in the guidelines for directing caisson sump
discharges to the outfall canal. Additionally, the proposed
revisions will modify Section VI to be consistent with the guidance
provided by NRC Draft Generic Letter for 10 CFR 20 Modification to
Technical Specifications (58 FR 68171, dated December 23, 1993).
These changes in effluent limits are not related to the probability
or consequences of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed revisions to the HBPP Technical Specifications are
administrative in nature and do not change the method by which any
safety-related system performs its function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed revisions to the HBPP Technical Specifications do
not affect the margin of safety associated with parameters for any
accident analysis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the analysis of the licensee and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Humboldt County Library, 636 F
Street, Eureka, California 95501.
Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas
& Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: Seymour H. Weiss.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California

Date of amendment request: November 23, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications Section VII.C, Plant Staff, to
decrease the minimum staff requirements for the shift operating
organization from five to two persons.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The probability or consequences of an accident previously
evaluated will not be affected by the change in plant staffing. The
plant staff manning requirements for the shift operating
organization are being reduced to reflect the condition of the plant
in a SAFSTOR mode. Previously evaluated accidents do not require
operator actions to mitigate or reduce the consequences of
occurrence. Consequently, the change will not affect the probability
or consequences of an accident occurring.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed revisions to the HBPP Technical Specifications are
administrative in nature. Further, there would not be any change in
equipment or system function or operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed revisions to the HBPP Technical Specifications do
not affect the margin of safety of any accident analysis since they
do not affect the parameters for any accident analysis, and they
have no effect on the current operating methodologies or actions
that govern plant performance.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the analysis of the licensee and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Humboldt County Library, 636 F
Street, Eureka, California 95501.
Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas
& Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: Seymour H. Weiss.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket No.
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County,
Pennsylvania

Date of application for amendment: January 13, 1995.
Description of amendment request: The proposed changes revise
Tables 3.7.1 and 3.7.4 to reflect a reduction in the number of primary
containment power operated outboard valves for the Traversing Incore
Probe (TIP) probes, and a redesignation of the containment penetration
numbers for the TIP ball, shear, and check valves. The proposed changes
are a result of PBAPS Modification P00068.
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 TIP system does not serve as an initiator or contributor to
any accidents previously evaluated. The system provides a means of
calibrating the Local Power Range Monitors and supports thermal
limit calculations. The new system performs the same function as the
old one. It will provide improved reliability and added redundancy
by allowing a complete flux mapping if a detector or drive failure
were to occur.
Installation of Modification P00068 and its operation will not
degrade any active or passive equipment that responds to an
accident. These changes do not decrease the
[[Page 11140]] effectiveness of equipment relied upon to mitigate
the previously evaluated accidents.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The modification is considered an enhancement to the TIP system
and does not serve as an initiator or contributor to any of the
accidents previously evaluated. The proposed changes do not
introduce a new mode of plant operation. The new system, like the
old one, is designed to keep the ball valves closed upon reset of
the Primary Containment Isolation System (PCIS) logic. The new TIP
control console will respond to a PCIS isolation signal in the same
manner as the old system.
Implementation of the proposed changes will not affect the
design function or configuration of any component or introduce any
new operating scenarios or failure modes or accident initiation.
Modification P00068 will not impair or prevent safety systems
from performing their safety function. It will not make any changes
to the design function of the TIP system. The classification of the
TIP ball and shear valves and their control circuitry will not
change as a result of this modification.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The TIP system does not serve as an initiator or contributor to
any accidents evaluated in the SAR [safety analysis report].
Modification P00068 is considered an enhancement to the existing TIP
system and does not change its design function. The reduction in the
number of containment penetrations from five to three does not
represent a reduction in a margin of safety because of additional
indexers in the new system. The proposed changes do not adversely
affect the assumptions or sequence of events used in any accident
analysis.
Therefore, the proposed changes do not involve a reduction in a
margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101.
NRC Project Director: John F. Stolz.

Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: June 13, 1994.
Description of amendment request: The proposed change would remove
license condition 2.E from the Facility Operating License. License
Condition 2.E incorporated the requirements of U.S. Department of
Interior publication ``Environmental Criteria for Electric Transmission
Systems''--1970, which applies to the construction cleanup,
restoration, and maintenance of transmission lines.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
(1) involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will remove a license condition unrelated to
nuclear safety. License condition 2.E incorporated into the
Operating License the requirements of U.S. Department of Interior
publication ``Environmental Criteria for Electric Transmission
Systems''--1970. The goal of this standard is to ``safeguard
aesthetic and environmental values within the constraints imposed by
the current state of high-voltage transmission technology.'' License
condition 2.E addresses the preservation of the environment and
natural resources. Removing this condition from the Facility
Operating License has no bearing on plant safety or the health and
safety of the public considering its non-nuclear nature. The
transmission line right-of-ways maintained by the [Power] Authority
[of the State of New York] are subject to regulation by other State
and Federal Agencies. Removal of this license condition will not
affect operation of safety related structures, systems or components
nor affect the quality assurance program at the FitzPatrick plant.
Therefore, the proposed change does 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.
License condition 2.E of the James A. FitzPatrick Plant
Operating License applies to the construction cleanup, restoration,
and maintenance of transmission lines. The Authority's transmission
lines are managed under guidelines based on the ``Generic
Transmission Line Right-of-Way Management'' plan requirements. The
requirements imposed by the plan on the FitzPatrick transmission
line right-of-ways exceed those of the U.S. Department of Interior
publication referenced in license condition 2.E in both scope and
details. Therefore, implementing the proposed change will 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.

License condition 2.E of the James A. FitzPatrick Plant
Operating License applies to the construction cleanup, restoration,
and maintenance of transmission lines. The requirements imposed by
this license condition are unrelated to nuclear safety. Continued
operation of the plant without Condition 2.E does not involve a
significant reduction in any margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: December 16, 1994; supplemented February
10, 1995 (TS 94-07).
Description of amendment request: The proposed change would reduce
the maximum allowed power levels and more clearly specify the plant
conditions allowed by the technical specifications for operation with
one or more main steam safety valves inoperable. In addition, the Bases
would be revised to reflect these changes and incorporate the revised
methodology used to establish the neutron flux setpoints.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of [[Page 11141]] Sequoyah Nuclear Plant (SQN)
in accordance with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change reduces the power level at which the reactor may be
operated with one or more main steam safety valves (MSSVs)
inoperable, to ensure that the secondary system is not
overpressurized during the most severe pressurization transient of
the secondary side. Additionally, this change will combine the TS
action statements for 3- and 4-loop operation with one or more MSSVs
inoperable, revise the mode requirements and times of Action
Statement 3.7.1.1.a, and correct a reference in the bases section to

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A95-4870. Public record. Not legal advice.