# Biweekly Notice

> Briefs, arguments, decisions, and more.

URL: https://www.frixlaw.com/law-library/documents/fr%3AX97-11022

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** October 22, 1997
- **Citation:** 62 FR 54866

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be

[[Page 54867]]

issued, under a new provision of section 189 of the Act. This provision
grants the Commission the authority to issue and make immediately
effective any amendment to an operating license upon a determination by
the Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 29, 1997, through October 9, 1997.
The last biweekly notice was published on October 8, 1997 (62 FR
52578).

Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By November 21, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the

[[Page 54868]]

Commission may issue the amendment and make it immediately effective,
notwithstanding the request for a hearing. Any hearing held would take
place after issuance of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Date of amendment request: April 7, 1997, as supplemented on August
7, 1997.
Description of amendment request: The proposed amendment would
revise the plants' technical specifications to permit replacement of
the 125 volt dc Gould batteries with new C&D Charter Power Systems,
Inc., batteries.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The replacement C&D battery has been selected to meet or exceed
the design, functional, and operational requirements of those of the
present Gould battery, including crosstie load limitations. The C&D
batteries are similar in design to the installed Gould batteries
(e.g., electrolyte specific gravity and construction of the plates)
except for capacity. The replacement C&D batteries have a
significantly larger capacity than the Gould batteries, which can
provide additional margin for future use. Also, the C&D batteries
are qualified for a 20 year life and meet the latest applicable
standards. The short circuit current provided by the C&D batteries
is well within the interrupting capability of the existing DC system
circuit breakers.
Additionally, the crosstie limit is increased to take advantage
of the larger C&D battery capacity. The C&D batteries were sized
based on having sufficient capacity to energize the design basis DC
loads for an operating unit with the IEEE-485 design margin while
maintaining the desired limited DC load of 200 amps for a shutdown
unit. This proposed change allows use of the C&D batteries' larger
capacity. The overall design, function, and operation of the DC
system and equipment has not been altered by these changes. The
proposed changes do not affect any accident initiators or precursors
and do not alter the design assumptions for the systems or
components used to mitigate the consequences of an accident as
analyzed in UFSAR Chapter 15. Therefore, there is no increase in the
probability or consequences of an accident previously evaluated.
B. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The replacement C&D batteries will provide the same functions as
those of the installed Gould batteries and will be operated with the
same types of operational controls. These limits include battery
float terminal voltage, individual cell voltage and electrolyte
specific gravity, and crosstie loading. Crosstie conditions are
allowed under the present Technical Specifications. The crosstie
limit is increased to take advantage of the larger C&D battery
capacity. The remaining changes are administrative in nature or
provide clarification to maintain consistency with other Technical
Specifications.
The DC system and its equipment will continue to perform the
same functions and be operated in the same fashion. The proposed
change does not create any new or common failure modes. The proposed
changes do not introduce any new accident initiators or precursors,
or any new design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, the possibility
of a new or different kind of accident from any accident previously
evaluated has not been created.
C. The proposed change does not involve a significant reduction
in a margin of safety.
The replacement C&D batteries will meet or exceed the design,
functional, and qualification requirements [of] those of the
installed Gould batteries. The proposed Technical Specification
limitations for the C&D batteries are derived from the same
methodology as the Gould batteries with applied margins in
accordance with IEEE-485. Increasing the crosstie loading limit
takes advantage of the larger C&D battery capacity with its
increased design margin. The proposed change to the crosstie loading
limit will continue to conservatively envelope the postulated design
requirements. The remaining changes are administrative in nature or
provide clarification to maintain consistency with other Technical
Specifications.
The inherent design conservatism of the DC system and its
equipment has not been altered. The DC system and its equipment will
continue to be operated with the same degree of conservatism.
Therefore, there is no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois

Date of amendment request: June 30, 1997, as supplemented on
September 25, 1997.
Description of amendment request: The proposed amendment would
revise the plants' technical specifications to permit the licensee to
take credit for soluble boron in spent fuel storage pool water to
maintain an acceptable margin of subcriticality.
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 following accidents have been specifically evaluated
relative to the SFP [spent fuel pool]: fuel assembly drop,
accidental misloading of spent fuel

[[Page 54869]]

assemblies into the SFP racks, and loss of normal cooling.
There is no increase in the probability of a fuel assembly drop
accident in the SFP when considering the presence of soluble boron
in the SFP water for criticality control. The handling of the fuel
assemblies in the SFP has previously been performed in borated
water. The criticality analysis shows the consequences of a fuel
assembly drop accident in the SFP are not affected when considering
the presence of soluble boron.
There is no increase in the probability of the accidental
misloading of spent fuel assemblies into the SFP racks when
considering the presence of soluble boron in the pool water for
criticality control. Fuel assembly placement will continue to be
controlled in accordance with approved fuel handling procedures and
the spent fuel storage configuration limitations. Periodic
surveillances of the SFP inventory (physical inventory and piece
counts) are performed in accordance with station procedures. These
surveillances ensure physical SFP inventory verification is
performed at least once per year and in a timely manner upon
completion of fuel movement in the SFP. The addition of credit for
decay time in the spent fuel pool in determining allowable storage
requirements is an extension of the reactivity equivalencing
methodologies used for burnup credit in WCAP-14416-NP-A,
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology,''
Revision 1, November 1996.
There is no increase in the consequences of the accidental
misloading of spent fuel assemblies into the SFP racks because
criticality analyses demonstrate that the pool will remain
subcritical following an accidental misloading if the pool contains
an adequate boron concentration. The proposed TS limitations and
surveillance frequency will ensure that an adequate SFP boron
concentration is maintained.
There is no increase in the probability of the loss of normal
cooling to the SFP water when considering the presence of soluble
boron in the pool water for subcriticality control since a high
concentration of soluble boron has previously been maintained in the
SFP water. A loss of normal cooling to the SFP water causes an
increase in the temperature of the water passing through the stored
fuel assemblies. This causes a decrease in water density which would
result in a decrease in reactivity when Boraflex neutron absorber
panels are present in the racks. However, since the proposed change
does not consider Boraflex to be present in the racks, and the SFP
water has a high concentration of boron, a density decrease causes a
positive reactivity addition. [The] consequences of this accident
are bounded by the misloaded assembly analysis. Because adequate
soluble boron will be maintained in the SFP water, the consequences
of a loss of normal cooling to the SFP will not be increased.
The proposed 48 hour surveillance frequency will be used to
verify the boron concentration is within the initial assumptions of
the criticality analysis. The current frequency of 24 hours was
based on the sampling frequency for reactor coolant system (RCS)
shutdown margin in Mode 5. A dilution of the SFP to a
keff greater than 0.95 would take a much longer time than
an RCS dilution resulting in loss of shutdown margin. This is due to
the larger SFP volume compared to the RCS volume, and the turnover
rate of water in the SFP is much less due to the lack of large
dilution sources for the SFP. The 48 hour sampling frequency is
sufficient based on operating experience, and based on the fact that
significant changes in the boron concentration in the spent SFP are
difficult to produce without detection, due to the large inventory
of water. Soluble boron concentration reduction requires the inflow
and outflow of large volumes of water which are readily detected by
SFP and fuel handling building sump high level alarms, flooding in
the fuel handling building or by normal operator rounds through the
SFP area (once every eight hours), allowing adequate time for
operator intervention prior to exceeding a keff of 0.95.
Therefore, consequences of an accident previously evaluated are not
increased by the change in surveillance frequency.
The format revisions to Specification 5.6.1.1 and reference to
the report containing the specific NRC-approved criticality
methodology in Specification 6.9.1.10 are administrative in nature
and will not result in an increase in the probability or
consequences of an accident previously evaluated.
Therefore, based on the above analysis, the proposed changes
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The results of criticality accident analyses in the SFP are
discussed in the UFSAR [Updated Final Safety Analysis Report] and in
Criticality Analysis Reports associated with previous licensing
activities. Specific accidents considered include fuel assembly
drop, accidental misloading of spent fuel assemblies into the SFP
racks, and loss of normal cooling.
LCO 3.9.1, ``BORON CONCENTRATION,'' contains limitations on the
boron concentration in the filled portions of the reactor coolant
system and the refueling canal during Mode 6. ComEd has maintained
soluble boron in the SFP at all times and has imposed administrative
limits on the SFP boron concentration, due in part to this
requirement. LCO 3.9.11 establishes specific boron concentration
requirements for the SFP water consistent with the results of the
new criticality analysis based on the NRC-approved methodology of
WCAP-14416-NP-A, ``Westinghouse Spent Fuel Rack Criticality Analysis
Methodology,'' Revision 1, November 1996. Credit is also taken for
radioactive decay time of the spent fuel.
Since soluble boron has always been maintained in the SFP water
and is currently controlled administratively, the implementation of
this requirement will have little effect on normal pool operations
and maintenance. The implementation of the proposed limitations on
the SFP boron concentration will only result in a requirement to
verify boron concentration of the SFP water every 48 hours rather
than every 24 hours. Sampling every 48 hours is sufficient to verify
the SFP boron concentration meets the assumptions of the criticality
analysis.
Because soluble boron has always been present in the SFP and has
been administratively controlled, a dilution of the SFP soluble
boron has always been a possibility. As shown in the SFP dilution
evaluation performed for Byron and Braidwood, a dilution of the SFP
which could increase the rack keff to greater than 0.95
(i.e., which could reduce the required margin to criticality) is not
a credible event.
Therefore, the implementation of the proposed limitations on the
SFP boron concentration and surveillance frequency will not result
in the possibility of a new kind of accident.
The proposed change to Specification 5.6.1.1 identifies the
requirements for the spent fuel rack storage configurations. The
proposed changes relate to the criteria for determining the storage
configuration. Since the proposed SFP storage configuration
limitations will be similar to those currently in the Byron and
Braidwood TS, these limitations will not have any significant effect
on normal SFP operations and maintenance and will not create any
possibility of a new or different kind of accident. Verifications
will continue to be performed to ensure that the SFP loading
configuration meets specified requirements.
The format revisions to Specification 5.6.1.1 and reference to
the report containing the specific NRC-approved criticality
methodology in Specification 6.9.1.10 are administrative in nature
and will not create the possibility of a new [or] different kind of
accident.
As discussed above, there is no significant change in plant
configuration or equipment and the proposed changes will not create
the possibility of a new or different kind of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed TS changes and the resulting spent fuel storage
operating limits will provide adequate safety margin to ensure that
the stored fuel assembly array will always remain subcritical. These
limits are based on a plant specific criticality analysis performed
in accordance with the NRC-approved Westinghouse spent fuel rack
criticality analysis methodology (WCAP-14416-NP-A). Credit is also
taken for radioactive decay time of the spent fuel.
Soluble boron credit provides significant negative reactivity in
the SFP such that the keff is maintained less than or
equal to 0.95. The proposed surveillance frequency will be used to
verify the boron concentration is within the initial assumptions of
the criticality analysis. A storage configuration has also been
defined, with a 95-percent probability at a 95-percent confidence
level, that ensures the spent fuel rack keff will be less
than 1.0 with no credit for soluble boron or Boraflex panels in the
racks. In addition to soluble boron credit, credit is taken for fuel
assembly burnup, decay time, and IFBAs [Integral Fuel Burnable
Absorber] when determining assembly storage requirements.

[[Page 54870]]

The loss of substantial amounts of soluble boron from the SFP
which could lead to exceeding a keff of 0.95 has been
evaluated and shown not to be credible. These evaluations show that
the dilution of the SFP boron concentration from 2000 ppm to 550 ppm
is not credible and that the spent fuel rack keff will
remain less than 1.0 when flooded with unborated water.
The format revisions to Specification 5.6.1.1 and reference to
the report containing the specific NRC-approved criticality
methodology in Specification 6.9.1.10 are administrative in nature
and will not result in a significant reduction in the plant's margin
of safety.
Therefore, the proposed changes in this license amendment will
not result in a significant reduction in the plant's 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: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
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: August 12, 1997
Description of amendment request: The proposed amendments would
remove a Technical Specification surveillance requirement to verify
that sediment deposition within the lake screenhouse is not greater
than one foot in thickness. Control of sediment accumulation in the
lake screenhouse would be accomplished through the Service Water
Performance Monitoring Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) Involve a significant increase in the probability or
consequences of an accident previously identified because:
Surveillance's [sic] to fully verify [that] the Ultimate Heat
Sink contains enough water to perform its design function will
continue. All cleanliness issues associated with ensuring
operability of Core Standby Cooling System - Equipment Cooling Water
System (CSCS-ECWS) equipment will be performed under the Service
Water Performance Monitoring Program, which meets GL 89-13
[Service Water System Problems Affecting Safety-Related
Equipment] recommended actions. By performing these
inspections per GL 89-13, LaSalle will ensure that there is no build
up of sediment, which could hinder or impede the design operation of
any safety or non-safety related equipment which takes a suction
from the service water tunnel. Based on the nature of sediment,
where it collects, and system design, the CSCS-ECWS will be
available if called upon or started to respond in case of an
accident for equipment cooling and long term cooling.
At no time, during approximately fourteen years of LaSalle
operation, has sediment built up or accumulated either in front of
the inlet to the CSCS cooling water screen bypass supply line or the
six 36-inch normal tunnel supply lines in such a manner that the
flow of water through these lines could have been reduced or
blocked. Instead, loose sediment collects in quiescent areas near
the traveling screens, the north end of the Service Water Tunnel,
under the outlets of the 36-inch normal tunnel supply lines in the
service water tunnel, and downstream of the butterfly isolation
valve in the 54 inch CSCS cooling water screen bypass supply line.
The sediment that collects in the service water tunnel does not
build up in a manner such that CSCS-ECWS, non-essential station
service water, or fire pump suctions from the tunnel are affected,
based on inspections since 1992.
The CSCS equipment cooling bypass valve, OE12-F300, is the
manual butterfly valve in the CSCS cooling water screen bypass
supply line. The bypass valve is being added to the ASME Section XI
Inservice Testing Program to cycle the valve quarterly. This valve
cycling will help maintain sediment level in the bypass line at a
low level due to flow through the line while the valve is not fully
closed and thus assure the bypass line remains available. The flow
is created due to the differential pressure across the circulating
water traveling screens with circulating water pumps in operation.
Therefore, neither essential nor non-essential service water
will be lost due to sediment. Neither the probability nor the
consequences of an accident are increased by the deletion of SR
4.7.1.3.c.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
Inspections for sedimentation will continue to be required by
LaSalle's Service Water System Performance Monitoring Program per GL
89-13, to ensure continued operability of Core Standby Cooling
System-Equipment Cooling Water System (CSCS-ECWS). The Ultimate Heat
Sink operability requires assurance of a specific volume of water to
provide cooling for at least 30 days for long term cooling following
an accident. The public will be protected by the safety analysis in
place by the fact that the safety and non-safety related equipment
which take a suction from the service water tunnel will not be
impaired by sediment. Therefore, there will be no possibility of a
new or different kind of accident from any accident previously
evaluated.
3) Involve a significant reduction in the margin of safety
because:
The Ultimate Heat Sink continues to be demonstrated Operable by
verifying a sufficient volume of water per TS SR 4.7.1.3.a and
4.7.1.3.b. Equipment operability will still be required per
Technical Specifications 3/4.7.1.1 and 3/4.7.1.2 for the CSCS-ECWS
systems. Sedimentation in the lake screenhouse is a maintenance/
cleanliness issue addressed by the LaSalle Service Water Performance
Monitoring Program. The program ensures equipment operability by
both inspection for and removal of sedimentation and chemical
control with a biocide to limit the growth of biological material
and silt dispersant to help keep silt in the flow stream from
coagulating. Therefore, there is minimal or no reduction in the
margin of safety due to the deletion of this surveillance
requirement.
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: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan

Date of amendment request: September 29, 1997 (NRC-97-0089)
Description of amendment request: The proposed amendment would
relocate the requirements for selected instrumentation and the
associated Bases from the technical specifications (TS) to the updated
final safety analysis report. The affected instrumentation is seismic
monitoring (TS 3.7.2), meteorological monitoring (TS 3.7.3), the
traversing in-core probe system (TS 3.7.7), the chlorine detection
system (TS 3.7.8), and the loose parts detection system (TS 3.7.10).
Changes to the TS index and list of tables were also requested to
reflect the relocation of these TS and associated Bases. NRC Generic
Letter 95-10, ``Relocation of Selected Technical Specification
Requirements Related to Instrumentation,'' dated December 15, 1995,
provided information concerning relocation of the requirements for
these instruments.

[[Page 54871]]

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 changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes would relocate TS 3/4.3.7.2 - Seismic
Monitoring Instrumentation, TS 3/4. 3.7.3 - Meteorological
Monitoring Instrumentation, TS 3/4.3.7.7 - Traversing In-Core Probe
System, TS 3/4.3.7.8 - Chlorine Detection System, and TS 3/4.3.7.10
- Loose-Part Detection System and their associated Bases to the
Fermi 2 Updated Final Safety Analysis Report (UFSAR). They would
also delete the special reporting requirements from the
aforementioned TS which contain such requirements. The proposed
changes would revise the TS Index and List of Tables to reflect the
relocation of these TS and associated Bases. The relocated TS
changes would be controlled in accordance with the requirements of
10 CFR 50.59.
The proposed changes affect TS that do not meet the NRC's
``Final Policy Statement on Technical Specification Improvements for
Nuclear Power Reactors'' or 10 CFR 50.36(c)(2)(ii) criteria for
inclusion in TS. These TS relocations are consistent with NUREG-
1433, ``Standard Technical Specifications, General Electric Plants,
BWR/4,'' Revision 1, April 1995. Furthermore, these five TS are
specifically identified in NRC Generic Letter 95-10, ``Relocation of
Selected Technical Specifications Requirements Related to
Instrumentation,'' dated December 15, 1995, as suitable for
relocation to licensee-controlled documents.
The Special Report requirements of TS 3/4.3.7.2, TS 3/4.3.7.3,
and TS 3/4.3.7.10 would be deleted as part of their relocation to
the UFSAR. The NRC reporting criteria of 10 CFR 50.72, ``Immediate
Notification Requirements for Operating Nuclear Reactors,'' and 10
CFR 50.73, ``Licensee Event Report Systems'' provide appropriate
requirements for reporting degraded and non-conforming conditions to
the NRC.
These proposed TS changes do not involve a significant increase
in the probability of an accident previously evaluated because no
changes are being made to any accident initiator. No previously
analyzed accident scenario is changed, and initiating conditions and
assumptions remain as previously analyzed.
These proposed TS changes do not involve a significant increase
in the consequences of an accident previously evaluated because the
proposed changes do not affect accident sequences or assumptions
used in evaluating the radiological consequences of an accident. The
proposed changes do not alter the source term, containment isolation
or allowable radiological releases.
2. The changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not change the way in which the plant is
operated and no new or different failure modes have been defined for
any plant system or component. No limiting single failure has been
identified as a result of the proposed changes. No new or different
types of failures or accident initiators are introduced by the
proposed changes.
3. The changes do not involve a significant reduction in the
margin of safety.
The proposed changes involve instrumentation and systems which
are not inputs in the calculation of any safety margin with regard
to Technical Specification Safety Limits, Limiting Safety System
Settings, Limiting Control Settings or Limiting Conditions for
Operation, or other previously defined margins for any structure,
system, or component.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226
NRC Project Director: John N. Hannon

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania

Date of amendment request: September 11, 1997
Description of amendment request: The proposed amendments would
relocate the reactor trip system and engineered safety feature
actuation system response times from technical specification (TS)
tables 3.3-2 and 3.3-5 to Section 3 of the licensee's Licensing
Requirements Manual (LRM) in accordance with the guidance provided in
NRC Generic Letter 93-08. Subsequent changes to the LRM would be
controlled in accordance with the requirements of 10 CFR 50.59. The
proposed amendments would also make several editorial changes in TSs
3.3.1.1 and 3.3.1.2, as well as making conforming changes to the Bases
for these TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment relocates the instrument response time
limits for the reactor trip system (RTS) and engineered safety
feature actuation system (ESFAS) from the technical specifications
to the Licensing Requirements Manual (LRM). The Core Operating
Limits Report (COLR) and containment penetrations table (containment
isolation valves) are controlled and maintained in the LRM. The LRM
was developed to control and maintain those items removed from the
technical specifications. The proposed amendment conforms to the
guidance given in Enclosures 1 and 2 of Generic Letter 93-08.
Neither the response time limits nor the surveillance requirements
for performing response time testing will be altered by this
submittal. The overall RTS and ESFAS functional capabilities will
not be changed and assurance that action requirements of the
protective and engineered safety features systems are completed
within the time limits assumed in the accident analyses is
unaffected by the proposed amendment. Therefore, operation of the
facility in accordance with the proposed amendment will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed amendment will not change the physical plant or the
modes of plant operation defined in the operating 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 will 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 measurement of instrumentation response times at the
frequencies specified in the technical specification provides
assurance that actions associated with the protective and engineered
safety features systems are accomplished within the time limits
assumed in the accident analyses. The response time limits, and the
measurement frequencies remain unchanged by the proposed amendment.
The proposed changes do not alter the basis for any other technical
specification that is related to the establishment of or maintenance
of a nuclear safety margin. Therefore, operation of the facility in
accordance with the proposed amendment will not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts &

[[Page 54872]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John F. Stolz

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: September 18, 1997
Description of amendment request: The amendment would decrease the
safety limit for the minimum critical power ratio (MCPR) from 1.12 to
1.11 for two recirculation loop operation and from 1.14 to 1.12 for
single recirculation loop operation in Technical Specification (TS)
2.1.1.2. Because the proposed amendment is for Cycle 10 operation, the
amendment would also revise the footnotes to TSs 2.1.1.2 and 5.6.5 to
state that the MCPR values and the items 19 and 20 are ``applicable
only for Cycle 10 operation.'' Cycle 10 operation is after the next
(i.e., 9th) refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The Minimum Critical Power Ratio (MCPR) safety limit is defined
in the Bases to Technical Specification [TS] 2.1.1 as that limit
which ``ensures that during normal operation and during Anticipated
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in
the core do not experience transition boiling.'' The MCPR safety
limit is re-evaluated for each reload and, for GGNS [Grand Gulf
Nuclear Station, Unit 1] Cycle 10, the analyses have concluded that
a two-loop MCPR safety limit of 1.11 based on the application of
GE's [General Electric Company's] cycle-specific MCPR safety limit
methodology is necessary to ensure that this acceptance criterion is
satisfied. For single-loop operation, a MCPR safety limit of 1.12
based on GE's cycle-specific MCPR safety limit methodology was
determined to be necessary. Core MCPR operating limits are developed
to support the Technical Specification [TS] 3.2 requirements and
ensure these safety limits are maintained in the event of the worst
case transient. Since the MCPR safety limit will be maintained at
all times, operation under the proposed changes will ensure [that]
at least 99.9% of the fuel rods in the core do not experience
transition boiling. Therefore, these changes to the [MCPR] safety
limit do not affect the probability or consequences of an accident
[previously evaluated].
GE's GESTAR-II approved methodology will continue to be
implemented and has no effect on the probability or consequences of
any accidents previously evaluated. One exception to GESTAR is that
the mis-oriented and mis-located bundle events will continue to be
analyzed as accidents subject to the acceptance criteria in the
current licensing basis [for GGNS]. The design of the GE11 fuel
bundles[, to be added to the core to replace Siemens fuel bundles,]
is such that the bundles are not likely to be mis-oriented or mis-
located and the normal administrative controls will be in effect for
assuring proper orientation and location. Therefore, the probability
of a fuel loading error is not increased. This analysis ensures that
postulated dose releases will not exceed a small fraction (10
percent) of 10CFR100 [10 CFR Part 100] limits. Therefore, the
probability or consequences of accidents previously evaluated are
unchanged.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The GE 11 fuel to be [added to the core and] used in Cycle 10
[operation] is of a design compatible with fuel present in the core
and used in the [current 9th] cycle. [The current core is a mixture
of GE11 and Siemens fuel bundles. The addition of GE11 to the core
for the 9th cycle is addressed in Amendment 131 to the license dated
November 21, 1996.] Therefore, the GE11 fuel will not create the
possibility of a new or different kind of accident. The proposed
changes do not involve any new modes of operation, any changes to
setpoints, or any plant modifications.
They introduce revised MCPR safety limits that have been proven
to be acceptable for Cycle 10 operation. Compliance with the
applicable criterion for incipient boiling transition continues to
be ensured. The proposed MCPR safety limits do not result in the
creation of any new precursors to an accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The MCPR safety limits have been evaluated in accordance with
GE's current cycle-specific methodology to ensure that during normal
operation and during AOOs, at least 99.9% of the fuel rods in the
core are not expected to experience transition boiling. Unless
otherwise approved, GGNS will implement only the NRC-approved
revisions to GE's GESTAR methodology. This GE methodology is similar
to those SPC [(Siemens Power Corporation)] reports current listed in
TS 5.6.5 and it will be applied in a similar, conservative fashion.
[TS 5.6.5, Core Operating Limits Report, lists the analytical
methods which are approved by NRC and are used to determine the core
operating limits for the GGNS core, including the MCPR.] One
exception to GESTAR is that the mis-oriented and mis-located bundle
events will continue to be analyzed as accidents subject to the
acceptance criteria in the current [GGNS] licensing basis. This
analysis ensures that postulated dose releases will not exceed a
small fraction (10 percent) of 10CFR100 limits. [The proposed
changes are to maintain the margin of safety for transition boiling
in the core.] On this basis, the implementation of this GE
methodology does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: James W. Clifford, Acting

Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: September 25, 1997
Description of amendment request: The proposed change modifies
Limiting Condition for Operation (LCO) 3.6.1.2 (Containment Leakage),
the associated Action, and Surveillance Requirement (SR) 4.6.1.2 in
Technical Specification (TS) for Waterford Steam Electric Station, Unit
3 (Waterford 3). The air lock door seal leakage rate acceptance
criteria in TS 6.15 is being changed from 0.01La to 0.005La. TS 6.15 is
also being modified to make the terms used in the Containment Leakage
Rate Testing Program consistent with terms used in the TS. This change
corrects an error that inadvertently decreased the allowed outage time
from 24 hours to 1 hour when the containment purge valve or containment
air lock leakage rates are not within limits. This error was made in
the Waterford 3 TS change request that was approved in Amendment 124
for Waterford 3 on April 10, 1997.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The proposed change adds the specific type of containment
leakage to the Limiting Condition for Operation (LCO), Action, and
Surveillance Requirement (SR) in the Containment Leakage Technical
Specification (TS) which results in increasing

[[Page 54873]]

the allowed outage time from 1 hour to 24 hours when the containment
purge valve or containment air lock leakage rates are not within
limits. The proposed change revises the air lock door seal leakage
rate acceptance criteria. Also, the proposed change revises the
Actions in the Containment Leakage TS to be consistent with the
Applicability, and revises terms in the Containment Section and
Administrative Controls Section of the TS to be consistent with the
Containment Leakage Rate Testing Program. This change will not
affect the probability of an accident. The containment purge valve
and air lock leakage rates are not an initiator of any analyzed
event. This change corrects two errors that were made in the
Waterford 3 10CFR50 Appendix J, Option B, TS change request that was
approved in TS Amendment 124. The first error inadvertently
decreased the allowed outage time from 24 hours to 1 hour when
either the containment purge valve or containment air lock leakage
rate acceptance criteria is not met. The second error inadvertently
increased the acceptance criteria for the air lock door seal
leakage. The revised air lock door seal leakage rate acceptance
criteria was never used at Waterford 3. This change also
administratively changes the Containment Leakage TS Action and terms
in the TS for consistency.
The proposed change will not affect the consequences of an
accident. The amount of leakage from the containment purge valve and
from the containment air lock will still be included in the overall
combined containment leak rate. Neither the overall containment
leakage rate limit nor the Action required to be taken if the
overall containment leakage rate were exceeded is being changed. The
Containment Leakage TS Action will be consistent with the
Applicability and TS 3.0.4 will prohibit entry into Mode 4 (RCS
[Reactor Coolant System] temperature 200 deg.F), unless
the overall containment leakage rate is within limit. The revised
air lock acceptance criteria was never used. Waterford 3 will
continue using the more restrictive acceptance criteria which is
controlled administratively. This proposed change does not affect
the mitigation capabilities of any component or system, nor does it
affect the assumptions relative to the mitigation of accidents or
transients.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No
The proposed change adds the specific type of containment
leakage to the LCO, Action, and SR in the Containment Leakage TS.
This results in increasing the allowed outage time from 1 hour to 24
hours when the containment purge valve or containment air lock
leakage rates are not within limits. The proposed change revises the
air lock door seal leakage rate acceptance criteria. Also, the
proposed change revises the Actions in the Containment Leakage TS to
be consistent with the Applicability, and revises terms in the
Containment Section and Administrative Controls Section of the TS to
be consistent with the Containment Leakage Rate Testing Program.
Neither the design nor configuration of the plant, or how the plant
is operated is being changed due to the addition of the specific
types of leakage from the Containment Leakage Rate Testing Program,
corrections made to the air lock door seal leakage rate acceptance
criteria, or the changes made to make the TS consistent. There has
been no physical change to plant systems, structures, or components
nor will these changes reduce the ability of any of the safety-
related equipment required to mitigate anticipated operational
occurrences or accidents. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No
The proposed change adds the specific type of containment
leakage to the LCO, Action, and SR in the Containment Leakage TS.
This results in increasing the allowed outage time from 1 hour to 24
hours when the containment purge valve or containment air lock
leakage rates are not within limits. The proposed change revises the
air lock door seal leakage rate acceptance criteria. Also, the
proposed change revises the Actions in the Containment Leakage TS to
be consistent with the Applicability, and revises terms in the
Containment Section and Administrative Controls Section of the TS to
be consistent with the Containment Leakage Rate Testing Program. The
proposed revision to the Action and making the containment leakage
rate terms consistent are administrative changes that have no
technical impact on the TS.
The pre-amendment 124 Waterford 3 TS and NUREG-1432 allowed
entry into specific Actions with allowed outage times greater than 1
hour (24 hours) when the air lock and purge valve leakage rate
acceptance criteria could not be met. This change restores this
allowed outage time which was inadvertently changed due to an error
in the TS change request. The increased allowed outage time may
prevent an unnecessary plant shutdown which is a plant transient.
Plant shutdowns produce thermal stress on components in the Reactor
Coolant System and the potential for a plant upset that could
challenge safety systems. This change decreases the possibility of a
plant shutdown by replacing the 1 hour allowed outage time with a 24
hour allowed outage time when the containment purge valve or
containment air lock leakage is not within limits. Also, the overall
containment leakage rate limits are not being changed and are
required to be maintained.
The revision to the air lock door seal acceptance criteria is a
more restrictive change to correct an error made by Waterford 3 in
the TS change request approved in Amendment 124. The less
restrictive acceptance criteria was never used; Waterford 3
continued testing to the more restrictive acceptance criteria.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: James W. Clifford, Acting

Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: October 1, 1997
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) for the Crystal River Nuclear
Electric Generating Plant Unit 3 (CR-3). The proposed TS change would
add a new TS section, 5.6.2.10.4.c. The new section will provide growth
monitoring criteria for the first span section of tubes in the ``B''
Once-Through Steam Generator (OTSG) with pit-like intergranular attack
(IGA) indications.
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
Does Not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
The purpose of OTSG tube inspection is to identify tubes that
have a higher potential for in service failure due to degradation
that results in a reduced ability to withstand normal and upset
operating conditions. The formal incorporation of specific
indication growth monitoring and repair criteria is consistent with
this purpose. Therefore, the probability of an accident previously
evaluated has not been increased.
Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR)
provides an analysis to assess the consequences of a steam generator
tube rupture event, including the complete severance of a steam
generator tube. This analyses concluded that CR-3 was sufficiently
designed to ensure that in the event of a steam generator tube
rupture, the radiological doses would not exceed the allowable
limits prescribed by 10 CFR 100. Neither would this result in
additional tube failures and further degradation of the

[[Page 54874]]

integrity of the reactor coolant pressure boundary. The proposed
changes do not alter this analysis in any fashion. Therefore, the
consequences of an accident have not been increased.
Criterion 2
Does not Create the Possibility of a New or Different Kind of
Accident from any Accident Previously Evaluated.
This change does not alter the design or operation of the OTSGs.
The incorporation of the proposed requirements is more conservative
than the existing ITS requirements. Neither the type of inspection
of OTSG tubes nor the process for performing inspections will be
changed by this amendment. Therefore, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Criterion 3
Does Not Involve a Significant Reduction in the Margin of Safety
as defined in the Bases for any Technical Specifications.
The previously performed analyses on the effects of OTSG tube
failures, as reported in the CR-3 FSAR, have demonstrated that
onsite and offsite consequences are within allowable limits. The
proposed change incorporates more conservative growth monitoring and
operational assessment criteria for the ``B'' OTSG first-span pit-
like IGA indications. This change does not result in a significant
reduction in the margin of safety as defined in the Bases for any
Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon

North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: September 26, 1997
Description of amendment request: The proposed amendment would
separate the requirements for Control Room Air Conditioning from
Control Room Makeup Air and Filtration as presently contained in
Technical Specification 3.7.6, ``Control Room Emergency Makeup Air and
Filtration,'' and its associated BASES. Technical Specification 3.7.6
now requires that each subsystem of Control Room Emergency Makeup Air
and Filtration include an OPERABLE emergency filtration unit and air
conditioning unit. The proposed amendment would separate the
requirements based on system function. The proposed amendment also
would increase the allowed outage time for the air conditioning portion
of the Control Room Air Conditioning Subsystem.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10
CFR 50.92(c)(1)). The proposed changes have no impact on the
probability of an accident because the control room ventilation
systems are support systems which have a role in the detection and
mitigation of accidents but do not contribute to the initiation of
any accident previously evaluated. Reorganizing the Technical
Specifications by function is merely an administrative change and
the change has no impact on the course of any accidents previously
evaluated since there is no change in the functions provided by the
subsystems.
Increasing the allowed outage time to 30 days from 7 days for
the cooling of recirculated air while one train is inoperable does
not affect the availability of the second train of air conditioning
or the actions required if both trains of air conditioning become
unavailable. Thus, the consequences accidents previously evaluated
are not increased.
B. The changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because they do not affect the function of any
facility structure, system or component, nor do they affect the
manner by which the facility is operated. The proposed changes do
not introduce any new failure modes.
C. The changes do not involve a significant reduction in a
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes
do not affect the function of any facility structure, system or
component, nor do they affect the manner by which the facility is
operated. Increasing the allowed outage time for the cooling of
recirculated air while one train is inoperable represents an
increase in the probability that the air conditioning functions
could be unavailable. However, the increase does not affect the
availability of the second train of air conditioning or the actions
required should both trains of air conditioning become unavailable.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833
Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270
NRC Project Director: Ronald B. Eaton, Acting

Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota

Date of amendment requests: September 26, 1997
Description of amendment requests: The proposed amendments would
revise Technical Specification (TS) 3.4.B, ``Auxiliary Feedwater
System,'' to provide specific guidance for conducting post-maintenance
operational testing of the turbine-driven auxiliary feedwater (TDAFW)
pump and associated system valves to meet operability and limiting
conditions for operation during unit startup. An additional change is
proposed to revise Table TS.3.5.2B to permit during Mode 2 the
bypassing of the auto start feature of the auxiliary feedwater (AFW)
pumps that results from the trip of both main feedwater pumps when the
feedwater pumps are not required to be operated.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Since none of the proposed changes involve a physical change to
the plant, the mechanisms that could cause a Loss of Normal
Feedwater have not changed. The probability that a Loss of Normal
Feedwater will occur is not altered.
This change still requires that the motor driven AFW Pump and
associated system valves are operable during Startup Operations.
Analysis of the Loss of Normal Feedwater transient shows that a
single AFW Pump provides sufficient AFW flow to prevent any adverse
conditions in the core. The condition of an inoperable TDAFW Pump is
already permitted during power operations where the consequences of
the event would be more severe than during startup. Since there are
no consequences from the Loss of Normal Feedwater event at power,
the consequences during startup would still be none, but the margins
would be larger because; (1) the amount of residual heat generated
is less because reactor power

[[Page 54875]]

at the start of the event is less and (2) the power history is lower
resulting in less decay heat.
Thus, these changes do not involve an increase in the
probability or consequences of an accident previously analyzed.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The proposed changes do not create the possibility of a new or
different kind of accident previously evaluated because the proposed
changes do not introduce a new mode of operation or testing, or make
physical changes to the plant.
The proposed changes do not alter the design, function,
operation, or testing of any plant component, therefore the
possibility of a new or different kind of accident from those
previously analyzed would not be created by these changes to
Technical Specifications.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
Margins previously established for the Loss of Normal Feedwater
event, were analyzed for different initial conditions. The Loss of
Normal Feedwater event was analyzed for Power Operations. This
analysis determined that no adverse conditions would occur in the
core. Since there are no consequences from the Loss of Normal
Feedwater event at power, the consequences during startup would
still be none but the margins would be greater because; (1) the
amount of residual heat generated is less because reactor power at
the start of the event is less and (2) the power history is lower
causing less decay heat.
Therefore, the proposed change does not result in a significant
reduction in the margin of safety currently established.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon

Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: September 3, 1997
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to revise the number of hours
operating personnel can work in a normal shift. The proposed amendment
also contains some administrative changes to the TSs.
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 proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
A. Establishing operating personnel work hours at, ``an 8 to 12
hour day, nominal 40 hour week,'' allows normal plant operations to
be managed more effectively and does not adversely effect
performance of operating personnel. Overtime remains controlled by
site administrative procedures in accordance with NRC Policy
Statement on working hours (Generic Letter 82-12). If 8 hour shifts
are maintained in part or whole, then acceptable levels of
performance from operating personnel is assured through effective
control of shift turnovers and plant activities. No physical plant
modifications are involved and none of the precursors of previously
evaluated accidents are affected. Therefore, this change will not
involve a significant increase in the probability or consequence of
an accident previously evaluated.
B. Editorial changes clarify section 6.2.2.g without changing
the intent or meaning. The proposed change meets the intent of the
NRC Policy Statement on working hours (Generic Letter 82-12).
C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the
intent or meaning of the technical specification sections.
Clarification to the table notation in section 4.1 related to the
definition of shift checks to monitor plant conditions will continue
as intended but are allowed to increase up to at least once per 12
hours. This increase is consistent with standard industry practice
as represented by the Standard Technical Specifications (STS),
Reference 1.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
A. Establishing operating personnel work hours at, ``an 8 to 12
hour day, nominal 40 hour week,'' allows normal plant operations to
be managed more effectively and does not adversely effect
performance of operating personnel. If 8 hour shifts are maintained
in part or whole, then acceptable levels of performance from
operating personnel is assured through effective control of shift
turnovers and plant activities. Overtime remains controlled by site
administrative procedures in accordance with the NRC Policy
Statement on working hours (Generic Letter 82-12). No physical
modification of the plant is involved. As such, the change does not
introduce any new failure modes or conditions that may create a new
or different accident. Therefore, operation in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any previously evaluated.
B. Editorial changes clarify section 6.2.2.g without changing
the intent or meaning. The proposed change meets the intent of the
NRC Policy Statement on working hours (Generic Letter 82-12).
C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the
intent or meaning of the technical specification sections.
Clarification to the table notation in section 4.1 related to the
definition of shift checks to monitor plant conditions will continue
as intended but are allowed to increase up to at least once per 12
hours. This increase is consistent with standard industry practice
as represented by the Standard Technical Specifications (STS),
Reference 1.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
A. Establishing operating personnel work hours at, ``an 8 to 12
hour day, nominal 40 hour week,'' allows normal plant operations to
be managed more effectively and does not adversely effect
performance of operating personnel. If 8 hour shifts are maintained
in part or whole, then acceptable levels of performance from
operating personnel is assured through effective control of shift
turnovers and plant activities. Overtime remains controlled by site
administrative procedures in accordance with the NRC Policy
Statement on working hours (Generic Letter 82-12) and is consistent
with the Standard Technical Specifications. The proposed change
involves no physical modification of the plant, or alterations to
any accident or transient analysis. There is no Basis to section 6
of the Technical Specifications, and the changes are administrative
in nature. Therefore, the change does not involve any significant
reduction in a margin of safety.
B. Editorial changes clarify section 6.2.2.g without changing
the intent or meaning. The proposed change meets the intent of the
NRC Policy Statement on working hours (Generic Letter 82-12).
C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the
intent or meaning of the technical specification sections.
Clarification to the table notation in section 4.1 related to the
definition of shift checks to monitor plant conditions will continue
as intended but are allowed to increase up to at least once per 12
hours. This increase is consistent with standard industry practice
as represented by the Standard Technical Specifications (STS),
Reference 1.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director

[[Page 54876]]

Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: September 8, 1997
Description of amendment request: The proposed amendment would
revise the f delta I function.
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 proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?

Response:
No. The revision to the negative [f delta I] penalty does not
significantly increase the probability or consequences of an
accident previously evaluated in the FSAR [Final Safety Analysis
Report]. This revision does not directly initiate an accident. The
consequences of accidents previously evaluated in the FSAR are
unaffected by this proposed change because no change to any
equipment response or accident mitigation scenario has resulted.
There are no additional challenges to fission product barrier
integrity.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
No. The revision to the negative [f delta I] penalty does not
create the possibility of a new or different kind of accident than
any accident already evaluated in the FSAR. No new accident
scenarios, failure mechanisms, or limiting single failures are
introduced as a result of this proposed change. The proposed
Technical Specification revision does not challenge the performance
or integrity of any safety related systems. Therefore, the
possibility of a new or different kind of accident is not created.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
No. The proposed change to the Technical Specification does not
involve a significant reduction in a margin of safety. The margin of
safety associated with the acceptance criteria for any accident is
unchanged.
The revision to the negative [f delta I] penalty will have no
affect on the availability, operability or performance of the safety
related systems and components and does not affect the plant
Technical Specification requirements. The revision to the negative
[f delta I] penalty does require a change to the Technical
Specifications but does not prevent inspections or surveillances
required by the Technical Specifications.
In addition, the revision to the [f delta I] parameters is based
upon the revised boron dilution rate used to analyze the boron
dilution transient. Indian Point 3 procedures require the placement
of one PW [primary water makeup] pump control switch in the pull-out
position, thus ensuring that only one PW pump is operating.
The Bases of the Technical Specifications are founded in part on
the ability of the regulatory criteria being satisfied assuming the
limiting conditions for operation for various systems. Conformance
to the regulatory criteria for operation with the revision to the
negative [f delta I] penalty is demonstrated and the regulatory
limits are not exceeded. Therefore, the margin of safety as defined
in the Technical Specifications is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601
Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New
York, New York 10019
NRC Project Director: S. Singh Bajwa, Director

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: September 29, 1997
Description of amendment request: The proposed amendment would
revise the Ginna Station Improved Technical Specifications (ITS) to
change the Allowable Value for high steam flow input into limiting
condition for operation (LCO) Table 3.3.2-1, Function 4.d.
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 Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. An increase in
the high steam flow Allowable Value for LCO Table 3.3.2-1, Function
4.d does not increase the probability of any analyzed accident nor
does it increase the likelihood of an inadvertent main steam
isolation. This function is not explicitly credited in the accident
analyses. Also, there are three coincident parameters which must be
reached in order for this function to cause a main steam line
isolation. It has been demonstrated that the change to the high
steam flow parameter does not delay the time at which this isolation
signal would be reached for any analyzed accident since the steam
flow value is reached much earlier in the accident scenario than the
other parameters. Therefore, these changes do not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or changes in
the methods governing normal plant operation. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes do not directly affect any analyzed
accident analysis. The new isolation times will not be affected for
analyzed accidents. As such, no question of safety is involved, and
the change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005
NRC Project Director: S. Singh Bajwa, Director

Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: August 26, 1997
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 3/4.6.1.3, ``Containment Systems -
Containment Air Locks,'' TS Bases 3/4.6.1.3, ``Containment Systems -
Containment Air Locks,'' and TS Bases 3/4.9.4, ``Refueling Operations -
Containment Penetrations.'' The containment air lock Limiting Condition
for Operation and Surveillance Requirements would be modified, and the
associated bases would be changed.
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

[[Page 54877]]

consideration, which is presented below:
The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power
Station, Unit No. 1, in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because accident initiators,
conditions, or assumptions are not affected by the proposed changes,
which clarify the Technical Specification (TS) Limiting Condition
for Operation (LCO) for the containment air locks, extend the test
frequency for the containment air lock interlock mechanisms, and
modify guidelines relative to the routing of hoses and cables
through the containment air lock during core alterations or during
movement of irradiated fuel within the containment.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
change the source term, containment isolation, or allowable
releases. The proposed changes do not affect the allowable
containment leakage rates presently specified in the Technical
Specifications.
The proposed change to Surveillance Requirement (SR) 4.6.1.3.c
to increase the surveillance interval for the air lock interlock
mechanism to ``at least once per REFUELING INTERVAL'' is justified
due to the purely mechanical nature of the interlock mechanism, and
given that the interlock mechanism is not normally challenged when
the air lock door is used for entry and exit since administrative
controls require strict adherence to single door opening. Operating
experience shows that the interlock mechanisms are very reliable.
Further, the proposed change will allow performance of the
surveillance under the conditions that apply during a plant outage,
which is preferable to performance, in part, with the plant at
power, as is currently necessitated by the present six month
interval surveillance requirement. Although an interlock mechanism
failure would not affect air lock sealing capabilities and would
therefore not directly affect containment integrity, performance of
the surveillance with the plant at power, when containment integrity
is required, carries with it the potential for loss of containment
integrity, should the interlock fail during testing and allow both
doors to be opened simultaneously. The proposed TS change may result
in an increased probability that due to the increased [decreased]
test frequency, an inoperable interlock mechanism could go
undetected for a longer length of time. However, in the unlikely
event that as a containment entry is being made, abnormal radiation
levels inside containment occur, any increase in consequences due to
a radioactive release as a result of an inadvertent opening of both
air lock doors (as could be allowed by a failed interlock mechanism
and assuming violation of administrative controls) is counter-
balanced by the decreased likelihood of similar events occurring
when the interlock mechanism is tested at power under the current,
more frequent, test requirement.
The proposed change to TS Bases 3/4.9.4 to add flexibility in
routing cable and hoses through the containment personnel air lock
will not affect the requirement to maintain at least one containment
personnel air lock door capable of being closed. The analysis
results for a fuel handling accident inside containment, as
presented in Section 15.4.7.3 of the DBNPS Updated Safety Analysis
Report (USAR), are well within the 10 CFR 100 guideline values.
Since the analysis does not take credit for containment isolation,
the status of the personnel air lock has no impact on the
acceptability of the results. Under the proposed change, in the
event of a fuel handling accident, release of radioactive material
will continue to be minimized since at least one personnel air lock
door will remain capable of being closed.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes. The proposed changes do not involve a change to the plant
design or operation and, therefore, will not introduce any new or
different failure modes or initiators.
3. Not involve a significant reduction in a margin of safety.
The proposed TS change to SR 4.6.1.3.c to increase the
surveillance interval for the air lock interlock mechanism will have
no adverse effect on plant safety based on its good historical
surveillance and maintenance data, and the reduction in testing at
power which will occur.
The analysis results for a fuel handling accident inside
containment, as presented in the D
Basis for proposed no significant hazards guideline values. Since
the analysis does not take credit for containment isolation, the status
of the personnel air lock has no impact on the acceptability of the
results. Therefore, the proposed change to TS Bases 3/4.9.4 to add
flexibility in routing cable and hoses through the containment
personnel air lock will not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: October 11, 1996
Description of amendment request: The proposed ammendment would
revise the Vermont Yankee Technical Specifications (TSs) regarding the
amount of foam concentrate required to support operability of the
Recirculation Motor Generator (M. G.) Set Foam System as stated in TS
3.13.G.1 and 3.13.G.2. In both instances, the required amount of foam
concentrate would be increased from 100 to 150 gallons.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated:
The changes proposed herein affect only the amount of foam
concentrate inventory required to support the operability of the
Recirculation M. G. Set Foam System and therefore does not modify or
add any initiating parameters that would significantly increase the
probability or consequences of any previously analyzed accident.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated:
These changes involve the upgrade of an existing system using
standard fire protection components to provide the level of
protection originally required. An evaluation has been completed to
ensure that the enhanced spray pattern and increased volume of spray
does not impact any equipment not previously evaluated and does not
create any threat of flooding to equipment. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety:
These changes do not affect any equipment involved in potential
initiating events or safety limits. Therefore, it is concluded that
the proposed change does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensees 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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301

[[Page 54878]]

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, N.W., Washington, DC 20037-1128
NRC Project Director: Ronald B. Eaton, Acting Director

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas

Date of amendment request: September 2, 1997
Description of amendment request: This license amendment request
proposes to revise Technical Specification 3.7.1.2, Auxiliary Feedwater
System, and associated Bases, to add requirements for the essential
service water (ESW) flowpaths to the turbine-driven auxiliary feedwater
pump (TDAFWP) and other changes consistent with the technical
specification conversion application previously submitted. The proposed
revisions would (a) provide an action and allowed outage time (AOT) for
inoperability of one of the redundant ESW flowpaths to the TDAFWP, and
(b) incorporate an action and AOT for inoperability of one of the
redundant steam flowpaths to the TDAFWP turbine and other changes to
make the auxiliary feedwater system limiting condition for operation
(LCO) and actions consistent with those previously submitted.
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.
ESW Flow Path Required Actions
This change would provide a 7-day AOT for the ESW supply flow
paths to the TDAFWP. This would replace administrative controls that
imposed a 72-hour AOT on ESW flow paths to the TDAFWP.
The proposed change does not result in any hardware changes or
changes to operating methodologies. This revision does not affect an
accident initiator of any analyzed accident since the TDAFWP ESW
supply only provides flow to equipment required to mitigate the
consequences of an accident. The revision recognizes that the TDAFWP
would remain available in most cases for accident mitigation because
of the low probability of an accident and subsequent equipment
failure requiring the use of the inoperable ESW supply for the
TDAFWP. Changing the AOT from 3 days to 7 days would have a
negligible effect on this small probability. Loss of the AFW
function would also require the failure of the MDAFWPs [motor-driven
auxiliary feedwater pumps]. In addition, the CST [condensate storage
tank] would be OPERABLE in accordance with LCO 3.7.1.3 and would be
available for use by the TDAFWP for all events except those external
hazards that represent a hazard to the integrity of the tank itself.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Steam Supply Flow Path Required Actions
This change would provide a 7-day AOT for the steam supply flow
paths to the TDAFWP. This would replace an administrative control
that required the TDAFWP to be declared inoperable without applying
an AOT. The proposed change does not result in any hardware changes
or changes to operating methodologies. This revision does not affect
an accident initiator of any analyzed accident since the TDAFWP
steam supply only provides power to equipment required to mitigate
the consequences of an accident. The revision recognizes the low
probability of an accident requiring the use of the inoperable steam
supply for the TDAFWP coincident with the failure of the MDAFWPs.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
3. Use of ``Trains'' Instead of ``Pumps and Associated Flow
Paths'' and Removal of Unnecessary Details
This change is partially administrative and partially a movement
of provisions not required to be in the technical specifications to
other controlled documents. The administrative change does not
impact initiators of analyzed events or equipment assumed in the
mitigation of accidents or transient events. The details moved from
the technical specification would be located in the Bases of the
technical specification. Since any changes to the Bases will be
evaluated per the requirements of 10 CFR 50.59, proper controls are
in place to adequately limit the probability or consequences of an
accident previously evaluated. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
4. Twelve Hours to HOT SHUTDOWN
This change would allow an additional 6 hours to achieve HOT
SHUTDOWN for the AFW System. The proposed change does not alter the
plant configuration or operation or function of any safety system.
Consequently, the change does not increase the probability of an
accident as defined in accident analysis. The proposed change
permits a longer time to cooldown to RHR [residual heat removal]
entry conditions; however, this would not affect the consequences of
any postulated accidents and is appropriate due to the need to avoid
any transients while cooling down with a potentially degraded AFW
System.
Therefore, the proposed change would have no significant effect
on the probability or consequences of any previously analyzed
accidents.
5. Additional AOT of 10 Days from Discovery of Failure to Meet
the LCO
The proposed change imposes more stringent requirements than
contained in current technical specification. The more stringent
requirements are imposed to ensure that the OPERABILITY requirements
for the AFW System are maintained consistent with the safety
analysis and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
6. Suspension of LCO 3.0.3
The proposed change involves clarifying the technical
specification. The proposed revision involves no technical changes
to the current technical specification. As such, this change is
administrative in nature and does not impact initiators of analyzed
events or assumed mitigation of accidents or transient events.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
1.ESW Flow Path Required Actions
The proposed change to add a 7-day AOT for the ESW supply flow
paths does not require physical alteration to any plant system or
change the method by which any safety-related system performs its
function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
2. Steam Supply Flow Path Required Actions
The proposed change to add a 7-day AOT for the steam supply flow
paths does not require physical alteration to any plant system or
change the method by which any safety-related system performs it
function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Use of ``Trains'' Instead of ``Pumps and Associated Flow
Paths'' and Moving of Unnecessary Details
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in controlling parameters. The proposed change will not
impose any different requirements and adequate control of the
information moved to the Bases will be maintained. The proposed
change will not impose any different requirements. Thus, this change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
4. Twelve Hours to HOT SHUTDOWN
The proposed change does not require physical alteration to any
plant system or change the method by which any safety-related system
performs its function. As discussed above, the change does allow
additional time to complete transfer from the SG [steam generator]
as the method for heat removal to the RHR System, but does not alter
the basic methodology.
Therefore, the proposed change would not create the possibility
of a new or different kind of accident.

[[Page 54879]]

5. Additional AOT of 10 Days from Discovery of Failure to Meet
the LCO
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in controlling parameters. The proposed change does
impose different (more restrictive) requirements. However, these
changes remain consistent with assumptions made in the safety
analysis regarding system OPERABILITY. Thus, this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
6. Suspension of LCO 3.0.3
The proposed change clarifies an implied requirement from
current technical specifications and does not involve a physical
alteration of the plant (no new or different type of equipment will
be installed) or changes in controlling parameters. The proposed
change will not impose any different requirements. Thus, this change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
1. ESW Flow Path Required Actions
The proposed change to add a 7-day AOT for the ESW flow paths
does not change any accident analysis assumptions, initial
conditions or results. Consequently, it does not have an effect on
margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
2. Steam Supply Flow Path Required Actions
The proposed change to add a 7-day AOT for the steam supply flow
paths does not change any accident analysis assumptions, initial
conditions or results. Consequently, it does not have an effect on
margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
3. Use of ``Trains'' Instead of ``Pumps and Associated Flow
Paths'' and Removal of Unnecessary Details
The proposed change will not reduce a margin of safety because
it has no impact on the design basis or safety analysis. In
addition, the requirements to be transposed from the technical
specification to the Bases are the same as the current technical
specification. Since any future changes to these requirements in the
Bases will be evaluated per the requirements of 10 CFR 50.59, proper
controls are in place to maintain an appropriate margin of safety.
Therefore, the changes do not involve a significant reduction in a
margin of safety.
4. Twelve Hours to HOT SHUTDOWN
The proposed change does not alter the basic regulatory
requirements or change any accident analysis assumptions, initial
conditions or results.
Therefore, the proposed change would have no significant adverse
effect on margins of safety.
5. Additional AOT of 10 Days from Discovery of Failure to Meet
the LCO
The imposition of more stringent requirements on AOT would
increase the margin of plant safety by providing additional
requirements to maintain AFW System OPERABILITY.
The change is consistent with the safety analysis and licensing
basis. Therefore, this change does not involve a reduction in a
margin of safety.
6. Suspension of LCO 3.0.3
The proposed change will not reduce a margin of safety because
it has no impact on the design basis or safety analysis. This change
is administrative in nature. As such, no question of safety is
involved.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman

Yankee Atomic Electric Company, Docket No. 50-029, Yankee Nuclear
Power Station, Franklin County, Massachusetts

Date of amendment request: September 5, 1997 (Accession No.
9709100106)
Description of amendment request: The proposed technical
specification (TS) changes are needed to permit removal of spent
nuclear fuel from the Spent Fuel Pit storage racks into a combined
storage/shipping cask and to enable handling of the cask components and
other hardware by the Yard Area Crane. Specific TS changes are needed
for minimum water coverage over spent fuel, shielding for personnel
exposure, increased loads carried over the fuel, addition of
restrictions for load paths over spent fuel and changes to the
appropriate TS bases.
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 changes provide for an alternate method of providing
protection of the spent fuel and spent fuel pit (SFP) from heavy
loads that must be transported over the SFP. The method chosen, that
is, providing a single-failure-proof overhead crane, is considered
an acceptable method as stated in Regulatory Guide 1.13, ``Spent
Fuel Storage Facility Design Basis,'' and NUREG-0612, ``Control of
Heavy Loads at Nuclear Power Plants.'' The Defueled Technical
Specification 3.1.2 requirement for five (5) feet of water above the
top of the fuel assemblies for fuel traveling in the SFP is provided
for personnel protection (ALARA). This protection is provided by the
shielding afforded by the shipping and/or transfer cask system. The
cask handling crane will comply with the single-failure-proof crane
design requirements of NUREG-0554, ``Single Failure-Proof Cranes for
Nuclear Power Plants,'' and meet the criteria specified in NUREG-
0612. In addition, design controls and administrative controls will
be maintained to prevent handling of the shipping and/or transfer
cask over spent fuel in the SFP. As such, these changes will not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated. NUREG-0612, Section
5, provides direction for providing an adequate level of defense-in-
depth for handling of heavy loads near spent fuel and safe shutdown
systems. The single-failure-proof overhead crane design is presented
as an acceptable method of providing the proper margin of safety for
handling of heavy loads. By upgrading the cask handling crane to a
single-failure-proof design and meeting the requirements presented
in Sections 5.1.1 and 5.1.6 of NUREG-0612 (for safe load path,
procedures, crane operator training and qualification, special
lifting devices, lifting devices that are not specially designed,
and crane inspection, testing, and maintenance) a sufficient level
of defense-in-depth is provided to ensure that a load drop is not a
credible event. As such, there is no increase in the probability or
consequence of an accident previously evaluated as a result of the
heavy load changes. A fuel handling incident is a currently analyzed
event; dropping of a fuel assembly over the spent fuel within the
transfer cask is similar to dropping of a fuel assembly over spent
fuel in the SFP. The design basis fuel handling event analysis
bounds these events, so there is no increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated. The defense-in-depth philosophy
provided by the single-failure-proof crane load handling sysem design,
and compliance with the requirements specified in Sections 5.1.1 and
5.1.6 of NUREG-0612 provide assurance that for a credible single
failure of the crane load handling system, the system will still be
able to perform its safety function. This provides assurance that a
load drop accident is not a credible event. As such, no new or
different kind of accident will be created from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety. The
proposed changes implement the guidelines of NUREG-0612 and
Regulatory Guide 1.13. YAEC is implementing an acceptable alternate
method of ensuring the safe handling of heavy loads

[[Page 54880]]

over the SFP. This method provides a defense-in-depth approach for
handling of heavy loads over the SFP and maintains the margin of
safety consistent with that of the current requirements. Further
protection is provided by the prohibition of these additional heavy
loads from travel over the spent fuel assemblies in the SFP racks.
The use of a single-failure-proof crane and associaed lifting
devices provide an increased margin of safety that ensure that a
load drop event is not credible and is considered an adequate
alternate for the additional area added to the safe load path. The
use of a limit switch to prevent movement of the prohibited cask
handling crane loads from movement beyond the safe load path,
provides an additional margin of safety, that was previously
provided by the steel framing at the southern edge of the SFP
superstructure roof opening. The single-failure-proof crane and
defense-in-depth design ensure that a load drop is not a credible
event, assuring that the margin of safety is not reduced.
Based on the above considerations, it is concluded that there is
reasonable assurance that the operation of Yankee Nuclear Power
Station consisent with the proposed changes will not endanger the
health and safety of the public.
The proposed change has been reviewed by the Plant Operations
Review Committee and the Nuclear Safety Audit and Review Committee.
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: Greenfield Community College, 1 College
Drive, Greenfield, Massachusetts 01301
Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One
International Place, Boston, Massachusetts 02110-2624
NRC Project Director: Seymour H. Weiss

Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: September 12, 1997
Brief description of amendment: The proposed amendment involves a
revision to the Emergency Diesel Generator protective relaying scheme
at CR3, as described in the Final Safety Analysis Report Chapter 8.
Date of publication of individual notice in the Federal Register:
September 30, 1997 (62 FR 51165).
Expiration date of individual notice: October 30, 1997
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.

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

Date of application for amendment: March 24, 1995, as supplemented
by letters dated September 10, 1995, and March 22, 1996.
Brief description of amendment: The amendment would change the
technical specifications (TS) to (1) reflect the applicable portions of
NUREG-1432, ``Standard Technical Specifications Combustion Engineering
Plants,'' (2) implement the recommendations of Generic Letter (GL) 93-
05, ``Line Item Technical Specification Improvements to Reduce
Surveillance Requirements for Testing During Plant Operation,'' and (3)
implement the recommendations of GL 94-01, ``Removal of Accelerated
Testing and Specific Reporting Requirements for Emergency Diesel
Generators.'' The purpose of the proposed amendment is to increase
emergency diesel generator (EDG) reliability by reducing stresses on
EDG caused by unnecessary testing. The associated Bases are also
updated.
Date of issuance: October 6, 1997
Effective date: October 6, 1997, to be implemented within 120 days
of date of issuance.
Amendment Nos.: Unit 1 - 114; Unit 2 - 107; Unit 3 - 86
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29870) The September 10, 1995, and March 22, 1996, supplemental letters
provided additional clarifying information and did not change the
original no significant hazards consideration. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
October 6, 1997. No significant hazards consideration comments
received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004

[[Page 54881]]

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

Date of application for amendments: March 28, 1996, as supplemented
November 20, 1996, and July 31, 1997.
Brief description of amendments: The amendments reduce the
moderator temperature coefficient limit shown on Technical
Specification Figure 3.1.1-1. This proposed change is necessary to
support changes in the safety analyses made to accommodate a larger
number of plugged steam generator tubes for future operating cycles.
Date of issuance: October 2, 1997
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 222 and 198
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Registe for amendment: February
21, 1997
Brief description of amendment: This amendment adds a specific time
limit to Technical Specification Table 3.3-3 to place an inoperable
refueling water storage tank level channel in a bypassed condition.
Date of issuance: September 30, 1997
Effective date: September 30, 1997
Amendment No.: 74
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17225) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

Date of application for amendments: July 1, 1997
Brief description of amendments: The amendments revise Technical
Specification Table 3.3.7.1-1, ``Radiation Monitoring
Instrumentation,'' to require two channels to be operable per trip
system as opposed to two per intake. This change reflects a
modification to the design of the instrumentation logic to satisfy
single failure requirements. The amendments also revise the associated
action statement to clarify system logic wording.
Date of issuance: October 9, 1997
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 121 and 106
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 27, 1997 (62 FR
45455). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 9, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois

Date of application for amendments: May 1, 1997
Brief description of amendments: The amendments clarify the load
value for the emergency diesel generator to be equal to or greater than
the largest single load and revise the frequency and voltage
requirements during the performance of the test.
Date of issuance: October 7, 1997
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 178 and 176
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33121). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 7, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021

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

Date of application for amendment: January 10, 1996, as
supplemented February 20, 1997
Brief description of amendment: The amendment revises the Technical
Specifications for the containment emergency escape air lock test
requirements. Concurrently, the Commission has also granted an
exemption to certain requirements of 10 CFR Part 50, Appendix J,
relating to the testing of the emergency escape air lock, to the extent
that leakage rate testing is not necessary after opening the emergency
escape air lock doors for post-test restoration or seal adjustment.
Date of issuance: September 30, 1997
Effective date: September 30, 1997
Amendment No.: 177
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 26, 1997 (62
FR 8795) The February 20, 1997, letter provided clarifying information
within the scope of the original application and did not change the NRC
staff's initial proposed no significant hazards considerations
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423 Consumers Energy Company, Docket No.
50-255, Palisades Plant, Van Buren County, Michigan
Date of application

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3AX97-11022. Public record. Not legal advice.