[Federal Register Volume 61, Number 244 (Wednesday, December 18, 1996)]
[Notices]
[Pages 66702-66721]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-31944]
[[Page 66702]]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 22, 1996, through December 6, 1996.
The last biweekly notice was published on December 4, 1996.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By January 17, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one
[[Page 66703]]
contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendment request: September 20, 1996.
Description of amendment request: The proposed amendments would
update the Pressure Temperature (P-T) curves contained in the Technical
Specifications to 22 Effective Full Power Years (EFPYs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
The proposed changes merely adjust the reference temperature for
the limiting beltline material to account for irradiation effects
and provide the same level of protection as previously evaluated.
The adjusted reference temperature calculations were performed
utilizing the guidance contained in Regulatory Guide 1.99, Revision
2. The change is administrative in nature to reflect the extension
of the operating limits to 22 EFPY. As such, these changes will not
significantly increase the probability or consequences of a
previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposed changes do not create the possibility of a new or
different kind of accident previously evaluated for Dresden or Quad
Cities Stations. No new modes of operation are introduced by the
proposed changes. The revised operating limits are merely an update
of the old limits by taking into account the effects of irradiation
on the limiting reactor vessel material. Use of the revised P-T
curves will continue to provide the same level of protection as was
previously reviewed and approved. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any previously evaluated.
The associated change to the P-T curves related to this proposed
amendment does not affect any activities or equipment and are not
assumed in any safety analysis to initiate any accident sequence for
Dresden or Quad Cities Stations; therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Involve a significant reduction in the margin of safety
because:
The proposed amendment reflects an update of the P-T curves to
extend the operating limit to 22 EFPY. The revised curves are based
on the latest NRC guidance along with actual data for the units. The
new limits retain the margin of safety to the level expected for a
new vessel, adjusted for irradiation effects as required by 10 CFR,
Appendix G, thereby maintaining a conservative margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
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: October 31, 1996.
Description of amendment request: The proposed amendments would
relocate the requirements for seismic monitoring instrumentation from
the Technical Specifications to licensee controlled documents. The
Technical Specifications affected are 3/4.3.7.2, ``Seismic Monitoring
Instrumentation,'' Table 3.3.7.2-1, ``Seismic Monitoring
Instrumentation,'' Table 4.3.7.2-1, ``Seismic Monitoring
Instrumentation Surveillance Requirements,'' and Bases Section 3/
4.3.7.2, ``Seismic Monitoring Instrumentation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 66704]]
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The function of the seismic monitoring instrumentation is to
monitor seismic activity above the Operating-Basis Earthquake (OBE)
threshold, and to record observed seismic data for comparison to
design basis response spectra. The seismic monitoring
instrumentation does not provide any function to mitigate an
accident or the consequences of an accident. The replacement seismic
monitoring instrumentation will remain in place. The proposed
Amendment is not a result of any changes to system function, alarm
setpoints, or main control room annunciators. Rather, the Technical
Specification requirements (as revised for the replacement
instrumentation) are being relocated to licensee-controlled
documents in accordance with NRC Generic Letter 95-10.
The proposed change relocates requirements and surveillances for
structures, systems, components or variables that do not meet the
criteria for inclusion in Technical Specifications as identified in
the Application of Selection Criteria to the LaSalle Technical
Specifications. The affected structures, systems, components or
variables are not assumed to be initiators of analyzed events and
are not assumed to mitigate accident or transient events. The
requirements and surveillances for these affected structures,
systems, components or variables will be relocated from the
Technical Specifications to an appropriate administratively
controlled document which will be maintained pursuant to 10 CFR
50.59. In addition, the affected structures, systems, components or
variables are addressed in existing surveillance procedures which
are also controlled by 10 CFR 50.59 and subject to the change
control provisions imposed by plant administrative procedures, which
endorse applicable regulations and standards. Therefore, this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The seismic monitoring instrumentation does not provide any
function to mitigate an accident or the consequences of an accident.
The replacement seismic monitoring instrumentation will remain in
place and will provide the same basic function as the existing
instrumentation. The replacement instrumentation will provide
enhanced system reliability and will not result in any changes to
system function, alarm setpoints, or main control room annunciators.
The Technical Specification requirements (as revised for the
replacement instrumentation) are being relocated to licensee-
controlled documents in accordance with NRC Generic Letter 95-10.
The proposed change does not involve any change in the methods
governing normal plant operation. The proposed change will not
impose or eliminate any requirements and adequate control of
existing requirements will be maintained. Thus, this change does not
create the 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 replacement seismic monitoring instrumentation will have no
impact on margin of safety. The intended function of the seismic
monitoring instrumentation, i.e. to record observed seismic data for
analysis to determine the impact on plant components, will be made
more reliable by this modification. The Technical Specification
requirements (as revised for the replacement instrumentation) are
being relocated to licensee-controlled documents in accordance with
NRC Generic Letter 95-10.
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumptions. In addition,
the relocated requirements and surveillances for the affected
structure, system, component or variable continue to meet the same
requirements as the existing Technical Specifications. However, the
LCO requirement specified in Section 3.3.7.2.a (to prepare and
submit a Special Report to the NRC within 10 days of the seismic
monitoring instrumentation being inoperable for more than 30 days)
will not be included in the ATR [Administrative Technical
Requirements] since the Technical Specification Special Report
requirements are only applicable to the LCOs. Since any future
changes to these requirements or the surveillance procedures will be
evaluated per the requirements of 10 CFR 50.59, no reduction in a
margin of safety will be permitted.
The existing requirement for NRC review and approval of
revisions, in accordance with 10 CFR 50.92, to these details
proposed for relocation does not have a specific margin of safety
upon which to evaluate. However, since the proposed change is
consistent with the BWR Standard Technical Specification, NUREG-
1434, Rev. 1 approved by the NRC Staff, revising the Technical
Specifications to reflect the approved level of detail ensures no
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: November 7, 1996.
Description of amendment request: The proposed amendments would
change Specification 4.3.1.A.4.b from verifying greater than or equal
to 17 percent steam generator secondary side wide range water level to
greater than or equal to 17 percent steam generator secondary side
narrow range water level.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of occurrence of any accident
previously evaluated.
Maintaining secondary side steam generator water level greater
than or equal to 17 percent by wide range level indication is the
current requirement by the technical specifications. By revising the
requirement to require using the narrow range water level, no change
in operating practices or plant configuration is made. The minimum
requirement of 17 percent by narrow range level indication is more
restrictive and conservative than 17 percent by wide range
indication. The requirement to maintain secondary side steam
generator water level greater than or equal to 17 percent by narrow
range indication is currently required by operations procedure PT-O,
Appendix F-1 and will be maintained. This change ensures that the
requirements for natural circulation cooldown are maintained in Mode
4. Therefore, changing this surveillance requirement does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not require a physical alteration of the
plant (no new or different equipment will be installed). The
Technical Specifications will continue to require OPERABLE steam
generator(s) for heat removal functions. The Technical
Specifications will continue to require the performance of SR
4.3.1.A.4.b. Changing the SR to use narrow level indication
correctly states the steam generator water level required to support
heat removal function. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes do not result in a significant reduction in
a margin of safety because it has no impact on any safety analysis
assumptions. The requirement to have OPERABLE steam generator(s) in
MODE 4 for heat removal function is maintained. The requirement to
perform SR 4.3.1.A.4.b is not changed. Changing the SR to use narrow
level indication correctly states the steam generator water level
required to support heat
[[Page 66705]]
removal function. Therefore, this 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
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. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: November 7, 1996.
Description of amendment request: The proposed amendments would
change the values for the reduced power range neutron flux high
setpoint trip that are specified when one or more code main steam
safety valves are inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of occurrence of any accident
previously evaluated.
The requirement to change the Power Range Neutron Flux High Trip
setpoints to the Reduced Setpoint Values of Table 3.7-1 for the most
restrictive loop if one or more code MSSVs are inoperable is not
changed by this amendment. As such, no change in operating practices
or plant configuration is being made.
The amendment provides new reduced setpoint values for the Power
Range Neutron flux High Trip to ensure that for the limiting
transient (Loss of Load/Turbine Trip [LOL/TT]), a secondary side
overpressurization condition does not occur. The new values were the
result of calculation using an algorithm provided by Westinghouse in
Westinghouse Nuclear Safety Advisory Letter NSAL-94-001, ``Operation
at Reduced Power Levels with Inoperable MSSVs,'' January 25, 1994.
The new values are much more restrictive than the previous values
and ensure that the probability or consequences of an accident
previously evaluated is not increased. Therefore, the new reduced
setpoint values do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not require a physical alteration of
the plant (no new or different equipment will be installed to
implement this change). The Reduced Neutron Flux High Trip setpoints
ensure that a secondary side overpressurization transient does not
occur for the most limiting transient. In addition, no new modes of
operations will be introduced by this change. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
This amendment provides new Reduced Power Range Neutron Flux
High Trip setpoints.The Specification that requires the Power Range
Neutron Flux High Trip setpoints be changed to the reduced values
for one or more inoperable MSSVs is not changed. The reduced Trip
setpoints are the result of new calculations using an algorithm
provided by Westinghouse in Westinghouse Nuclear Safety Advisory
Letter NSAL-94-001, ``Operation at Reduced Power levels with
Inoperable MSSVs,'' January 25, 1994, and ensure the LOL/TT
transient does not result in a secondary overpressurization.
Therefore, this 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
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. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: November 7, 1996.
Description of amendment request: The proposed amendments would
clarify the operability requirements for the residual heat removal
(RHR) loops during core alteration operations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of occurrence of any accident
previously evaluated.
The ability to remove an RHR loop from operation for up to one
hour per eight-hour period is currently allowed by technical
specification 3.13.9.B.b. By adding a reference to LCO [Limiting
Condition for Operation] 3.13.1.A.4. and adding the requirement to
suspend CORE ALTERATIONS to Action 3.13.9.B.a. to be consistent with
3.13.9.B.b., no change in operating practices or plant configuration
is made. By maintaining the requirement to have an RHR loop in
operation during MODE 6, and by requiring CORE ALTERATIONS to be
suspended if an RHR loop is not back in operation after one hour,
adequate corrective actions are implemented until the RHR loop is
restored to operating status. Therefore, operation of the system is
consistent with current technical specifications and this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not require a physical alteration of the
plant (no new or different equipment will be installed to implement
this change). The Technical Specifications will continue to require
an RHR loop to be in operation during MODE 6, and will only permit
the loop to be not in operation for up to one hour in an eight-hour
period. The Technical Specifications will continue to require
compliance with these limitations and suspension of CORE ALTERATIONS
if an RHR loop is not in operation for more than one hour. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes do not result in a significant reduction in
a margin of safety because it has no impact on any safety analysis
assumptions. The requirement to have an RHR loop in operation during
MODE 6 is maintained, along with the ability to remove RHR from
operation for up to one hour per eight-hour period. If an RHR loop
is not in service beyond 1 hour per TS 3.13.9.B, CORE ALTERATIONS
will be suspended. Therefore, this 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One
[[Page 66706]]
First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of amendment request: November 7, 1996.
Description of amendment request: The proposed amendment would
revise Technical Specification 4.2.9, Service and Instrument Air
System, to add an additional air compressor.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Utilizing the existing piping configuration, both the new and
the existing air compressors are capable of supporting either
portion of the Service and Instrument Air System. The addition of
the fourth air compressor will decrease the probability of an
accident previously evaluated, because capacity is being added to
the system. The consequences of an accident previously evaluated
will not be affected by the addition of a fourth air compressor. The
Service and Instrument Air System performs the non-safety related
function of providing compressed air for service use and moisture
free compressed instrument air for control air demands. The
instrument air portion is designed so that its operation is required
for plant reliability, not plant nuclear safety. Safety-related
equipment supplied by instrument air is designed to fail in its safe
condition upon loss of instrument air or, safety-related equipment
(and nonsafety-related equipment determined to be important to
safety) required to operate subsequent to instrument air failure is
supplied by backup nitrogen accumulators.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The operation of the equipment in the Service and Instrument Air
System is essentially unchanged. The new air compressor is a similar
design (nonlubricated), providing additional air volume at a quality
comparable to the three existing air compressors. Therefore, the
possibility of an accident of a different kind than any previously
evaluated has not been created.
(3) Involve a significant reduction in a margin of safety
The Technical Specification does not specify a margin of safety
for the operation of the Service and Instrument Air System, other
than specifying that [``Instrument and service] air shall be
supplied by three, nonlubricated air compressors, each rated at 70
scfm [standard cubic feet per minute]. Instrument air shall also
pass through a dryer.'' Addition of a fourth air compressor will
increase the available capacity, thus increasing the margin of
safety. Therefore, adding the statement ``and one, nonlubricated air
compressor rated at 100 scfm'' to Technical Specification 4.2.9.
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: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
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: October 4, 1996.
Description of amendment request: The proposed amendment would
revise the surveillance requirements in Technical Specifications (TSs)
4.1.2.3.1, 4.1.2.4.1, 4.5.2.b, and 4.6.2.1.b and associated Bases. The
subject surveillance requirements are applicable to the charging/high
head safety injection pumps, low head safety injection pump, and the
containment quench spray pumps. The proposed changes would replace the
current specific test acceptance criteria contained in these
surveillance requirements with requirements to verify pump performance
in accordance with the Inservice Testing Program, the Emergency Core
Cooling System Flow Analysis, or the Containment Integrity Safety
Analysis, as applicable. The proposed changes would also make minor
editorial changes in these TSs and make conforming changes in the TS
Index pages.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The change does not result in a modification to plant equipment
nor does it affect the manner in which the plant is operated. Since
the physical plant equipment and operating practices are not
changed, as noted above, there is no change in the probability of an
accident previously evaluated.
The proposed change will not lower the pump performance
operability criteria for the charging/high head safety injection,
low head safety injection and quench spray pumps, as assumed in the
safety analysis. The required values for developed pump head and
flow will continue to satisfy accident mitigation requirements and
will be maintained and controlled in the Inservice Testing (IST)
Programs(s).
Since the proposed change does not lower the pump's performance
acceptance criteria, as assumed in the safety analysis, the
containment depressurization system will continue to meet its design
basis requirements. The proposed change will not impose additional
challenges to the containment structure in terms of peak pressure.
The calculated offsite dose consequences of a design basis accident
(DBA) will remain unchanged since the one hour release duration and
source term remain unchanged. The ability of the emergency core
cooling system (ECCS) subsystems to provide sufficient emergency
core cooling capability in the event of a loss of coolant accident
(LOCA) remains unchanged. Therefore, peak cladding temperatures
during a LOCA will continue to remain within acceptable limits. The
ability of the ECCS subsystems to provide sufficient long term core
cooling capability in the recirculation mode during the accident
recovery period remains unchanged. The charging pumps, as part of
the boron injection system, will continue to provide sufficient flow
to ensure negative reactivity control during each mode of facility
operation. Future changes to the pump head and flow requirements
will be made under the 10 CFR 50.59 process to ensure that the
system performance requirements continue to be met.
The proposed change to the Bases section will ensure that safety
analyses assumptions for assumed pump performance continue to be
met. The words ``required developed head'' will be clearly defined
to reflect that they refer to the value(s) assumed in the safety
analysis for the pump's developed head at a specific or a given
point. The proposed changes to the Index pages and the footnote in
LCO 3.1.2.4 are administrative in nature and do not affect plant
safety.
Based on the above discussion, it is concluded that this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not alter the method of operating the
plant. The charging pumps will continue to be in service during
plant operation and be available to perform their function as high
head safety injection pumps. This proposed change does not pose
additional challenges to the design or function of the charging
pumps. The low head safety injection and quench spray
[[Page 66707]]
systems are accident mitigation systems and are normally in standby.
System operation would be initiated as required to mitigate the
consequences of a DBA. The charging/high head safety injection, low
head safety injection and quench pumps will continue to provide
sufficient flow to mitigate the consequences of a DBA. These
systems' operation continues [sic] [continue] to fulfill the safety
functions for which they were designed and no changes to plant
equipment will occur. As a result, an accident which is new or
different than any already evaluated in the Updated Final Safety
Analysis Report will not be created due to this change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety? The surveillance requirements for demonstrating that the
pumps are operable will continue to assure the ability of the system
to satisfy its design function. Therefore, the proposed change will
not affect the ability of these systems to perform their safety
function.
The containment systems' design requirements to restore the
containment to subatmospheric condition within one hour will
continue to be satisfied. This proposed change does not have an
effect on the containment peak pressure since the charging/high head
safety injection, low head safety injection and quench spray pumps'
performance requirements are not being lowered. The ability of the
ECCS subsystems to provide sufficient emergency core cooling
capability in the event of a LOCA remains unchanged. Therefore, peak
cladding temperatures during a LOCA will continue to remain within
acceptable limits. The ability of the ECCS subsystems to provide
sufficient long term core cooling capability in the recirculation
mode during the accident recovery period remains unchanged. The
charging pumps, as part of the boron injection system, will continue
to provide sufficient flow to ensure negative reactivity control
during each mode of facility operation. There is no resultant change
in dose consequences since source term remains unchanged and the
containment will continue to reach a subatmospheric pressure within
the first hour following a DBA.
Each pump's performance requirements will continue to be
controlled in a manner to ensure safety analysis assumptions are
met.
Therefore, based on the above discussions, it can be concluded
that the proposed change does not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant Units 1 and 2, St. Lucie County, Florida
Date of amendment request: October 31, 1996.
Description of amendment request: The proposed amendments will
revise administrative controls Technical Specification (TS) 6.5.1,
``Facility Review Group (FRG),'' and TS 6.8, ``Procedures and
Programs.'' The revisions to TS 6.5.1 reduce the scope of procedures
and procedure changes which require review by the FRG, transfer
approval of certain procedures from the Plant Manager to the FRG, and
require copies of FRG meeting minutes be provided to the Plant Manager.
The changes to TS 6.8 reflect the corresponding changes in TS 6.5.1,
and expand the scope of the section on temporary changes to procedures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments revise certain administrative controls
involved with the on-site programmatic process for review and
approval of plant procedures. Specifications that are in place to
provide assurance that the unit operating staff qualifications are
acceptable, and that written procedures are established, implemented
and maintained for safety related activities are not being changed.
The revisions are consistent with industry standards established
pursuant to 10 CFR Part 50, Appendix B, and do not alter any
parameter or equipment performance assumptions that are contained in
plant safety analyses to evaluate the initiation or consequences of
an accident. Therefore, operation of either facility in accordance
with its proposed amendment would not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments will not change the physical plant or
the modes of plant operation defined in the Facility License for
either St. Lucie unit. Changes proposed for the administrative
controls do not involve the addition or modification of equipment
nor do they alter the design or operation of plant systems.
Therefore, operation of either facility in accordance with its
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendments revise certain administrative controls
involving the on-site programmatic process for review and approval
of plant procedures. The scope, or the requirement to establish,
maintain, and implement procedures for activities that could affect
nuclear safety are not being changed. The proposed changes are
consistent with approved industry standards and do not alter the
basis for any technical specification that is related to the
establishment of, or the maintenance of, a nuclear safety margin.
Therefore, operation of either facility in accordance with its
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light,
11770 US Highway 1, North Palm Beach, Fl 33408.
NRC Project Director: Frederick J. Hebdon.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: October 31, 1996 (TSCR 205).
Description of amendment request: The proposed change requests
deletion of Technical Specification Table 3.5.2 which lists automatic
primary containment isolation valves. In addition, this change request
clarifies the applicability of an action statement which applies to
several limiting conditions for operation in Section 3.5 and deletes
closure time requirements for several automatic isolation valves in
Section 4.5.F.
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:
[[Page 66708]]
1. The proposed deletion of the automatic primary containment
isolation valve Table 3.5.2 and closure times for several valves in
Specification 4.5.F.1 are administrative in nature and do not affect
the purpose, function, operability and testing requirements of the
automatic primary containment isolation valves or the isolation
condenser isolation valves. The required action contained in
Specification 3.5.A.7 has been moved to the associated
specifications and has not changed. Capitalizing definitions and
deleting unneeded pages are also administrative changes which
enhance the usability of the Technical Specifications. Therefore,
the proposed changes do not increase the probability of occurrence
or consequence of an accident previously evaluated.
2. The proposed changes are administrative and do not involve a
physical change to plant configuration nor do they affect the
performance of any equipment. Existing limiting conditions for
operation and surveillance requirements are retained. Therefore, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
3. Deleting the list of valves in Table 3.5.2 and valve closure
times in Specification 4.5.F.1 are administrative changes which do
not affect the purpose or function of the automatic primary
containment isolation valves. The listing of the automatic primary
containment isolation valves and stroke time requirements will be in
controlled plant procedures. Changes to the list or closure times
can be made in accordance with review procedures required by Section
6.5 of the Technical Specifications and 10 CFR 50.59. Similarly,
inserting the statement of required action in Specification 3.5.A.7
into the Specifications to which it applies does not modify the
condition or the action to be taken and is an administrative change
which clarifies the Technical Specifications. Therefore, the margin
of safety is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: November 12, 1996, as supplemented
November 27, 1996 (TSCR 224).
Description of amendment request: The proposed technical
specification change will reflect the implementation of the revised 10
CFR Part 20, ``Standards for Protection Against Radiation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated.
The proposed revisions to the liquid release rate limits and
bases and gaseous effluent bases will not result in a change in the
types or amounts of effluents released nor will there be an increase
in individual or cumulative radiation exposures. In addition, these
changes do not impact the operation or design of any plant
structures, systems, or components. These changes ensure compliance
with 10 CFR 50.36a and 10 CFR 50 Appendix I and result in levels of
radioactive materials in effluents being maintained ALARA [as low as
is reasonably achievable]. The revision to the high radiation area
controls and dose measurement distance will ensure areas are
conservatively posted as high radiation areas in compliance with 10
CFR 20.1601(a)(1) and provide controls to ensure individuals are not
overexposed. Other proposed changes consist of revisions to 10 CFR
20 references to recognize the new section numbers, and
administrative controls for record keeping to maintain compliance
with the new Part 20.
These changes will not result in a change to plant design or
operation. Therefore, it can be concluded that the proposed changes
do not involve an increase in the probability or consequences of an
accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The
proposed changes do not affect the plant design or operation nor do
they result in a change to the configuration of any equipment. There
will be no change in the types or increase in the amount of
effluents released offsite.
Therefore, this proposed change cannot create the possibility of
a new or different kind of accident from any previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed revisions do not involve any change in the types or
increase in the amount of effluents released offsite. The proposed
changes do not involve any actual change in the methodology used in
the control of radioactive wastes or radiological environmental
monitoring. The methodology that will be used in the control of
radioactive effluents and calculation of effluent monitor setpoints
will result in the same effluent release rate as the current
methodology now being used. The operational flexibility needed for
releases allows the use of limits as proposed. In addition, the
changes in measurement distances for determination of high radiation
areas will not result in an increase in individual or cumulative
occupational radiation exposures since it will result in a more
conservative identification of high radiation areas. Compliance with
the limits of the new 10 CFR 20.1301 will be demonstrated by
operating within the limits of 10 CFR 50 Appendix I and 40 CFR 190.
Thus, operation of the facility in accordance with the proposed
amendment does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: August 29, 1996, as supplemented October
3, 1996. The October 3, 1996, submittal contained editorial changes
only and did not change the initial no significant hazards
consideration evaluation.
Description of amendment request: The purpose of this amendment
request is to incorporate certain improvements from the Standard
Technical Specifications for B&W Plants, NUREG-1430.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
GPU Nuclear has determined that this Technical Specification Change
Request involves no significant hazards consideration as defined in 10
CFR 50.92 because:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the
[[Page 66709]]
consequences of an accident previously evaluated. The proposed
amendment deletes limiting conditions for operation (LCOs) from the
TMI-1 Technical Specifications that are no longer required to be
addressed in Technical Specifications per 10 CFR 50.36(c)(2)(ii).
The proposed amendment deletes Surveillance Requirements from the
TMI-1 Technical Specifications that are related to the LCOs to be
deleted. These items are addressed in licensee controlled documents.
Certain design feature specifications are also to be deleted
consistent with the RSTS [Revised Standard Technical Specifications]
for B&W plants. The proposed changes do not modify the operation,
limits or controls of systems, structures or components relied upon
to prevent or mitigate the consequences or accidents previously
evaluated.
Also, the reliability of systems and components relied upon to
prevent or mitigate the consequences of accidents previously
evaluated is not degraded by the proposed changes. Therefore, this
change does not involve a significant increase in the probability of
occurrence or the consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated because no
new failure modes are created by the proposed changes.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety because the proposed amendment does not change any operating
limits for reactor operation.
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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: September 20, 1996.
Description of amendment request: The proposed amendment would
revise the Nine Mile Point Unit 1 (NMP1) Technical Specifications that
involve the frequencies of surveillance requirements stated in Tables
4.6.2a, 4.6.2b, 4.6.2g, and 4.6.11, and Sections 4.2.5b(1), 4.3.2b,
4.3.6b(1), 4.3.6b(2), 4.3.6b(3), 4.3.6b(4), 4.3.6c(2), 4.6.13b.1, and
4.6.13b.2. The surveillances associated with these tables and sections
are currently satisfied during NMP1 refueling outages prior to restart
of the unit. The proposed changes would permit surveillance testing
either while the reactor is operating or during outage periods not
associated with refueling. The requirements of the surveillance
sections and tables addressed by this request that are not changed to
be performed at power are being changed to allow surveillance credit to
be taken for performance of the associated surveillances while the
plant is in the Cold Shutdown, Refueling, or Major Maintenance modes.
In addition to these proposed changes, typographical errors are
corrected.
Basis for proposed no significant hazards consideration
determination: The licensee states that: ``The periods between
surveillances will not be inappropriately lengthened. For the affected
surveillances, NMP1 administrative controls will require that the
interval between surveillance testing not exceed a period equal to 1.25
times the nominal 24 months frequency (no longer than 30 months). The
NMP1 plant preventive maintenance and surveillance database will be
revised accordingly.''
The licensee groups the systems affected by this request into four
categories:
Category 1: The associated system will remain operable and able
to automatically perform its safety function during performance of
surveillances that satisfy the proposed surveillance requirement.
Category 2: The system is required for monitoring purposes only
and provides no automatic safety actuation function and redundant,
or redundant and alternate channels are available for required
monitoring.
Category 3: There is no change in the system configuration or
plant operating conditions during the performance of associated
surveillances whether the plant is shutdown for refueling or
shutdown for maintenance. The surveillances performed to meet the
requirements of NMP1 Technical Specifications Tables 4.6.2a
Parameter 8 and 4.6.2g Parameter 6 are included in this category and
may also be completed in concurrence with a unit shutdown. The only
difference between the proposed changes and the normal unit shutdown
sequence is that the mode switch may be taken to ``Shutdown'' in
order to scram the plant. The response of the plant is the same as
it is under the current plant shutdown procedures. There are no
other differences in testing techniques or testing criteria from
those previously required by the NMP1 Technical Specifications.
Category 4: The system or equipment is isolated or out of
service during the performance of the required surveillances. The
associated surveillance may be performed concurrently with quarterly
valve stroking, at which time the system or equipment is already out
of service.
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequence of an accident previously evaluated.
Each of the four categories [* * *] are evaluated separately below:
Category 1: The associated systems will remain operable and able
to automatically fulfill as designed any required safety functions
that may become necessary during performance of required
surveillances. No physical change to the plant design, materials, or
standards is involved. No change to instrumentation operating
characteristics outside current tolerances will be made. No plant
transients will be initiated as a result of the proposed changes. No
initiator of any accident previously evaluated is adversely
affected. No system required to actuate to respond to any accident
previously evaluated in the UFSAR [Updated Final Safety Analysis
Report] is adversely affected by the proposed change.
Category 2: The associated systems will be required for
monitoring purposes only and provide no automatic safety actuation
function and redundant, or redundant and alternate channels are
available for required monitoring. Since redundant monitoring
instrumentation will still be available as required by the technical
specifications, the associated systems' functions in accident
mitigation are not affected. No physical change to the plant design,
materials, or standards is involved. No change to instrumentation
operating characteristics outside current tolerances will be made.
No plant transients will be initiated as a result of the proposed
changes. No initiator of any accident previously evaluated is
adversely affected. No system required to actuate to respond to any
accident previously evaluated in the UFSAR is adversely affected by
the proposed changes.
Category 3: There will be no change in the system configuration
or plant operating conditions during the performance of associated
surveillances. The associated system's ability to perform required
safety functions will not be affected, whether the plant is shutdown
for refueling or shutdown for maintenance. The surveillances
performed to meet the requirements of NMP1 Technical Specifications
Tables 4.6.2a Parameter B and 4.6.2g Parameter 6 are included in
this category and may also be performed in concurrence with a unit
shutdown. The only difference between the proposed changes and the
normal unit shutdown sequence is that the mode switch
[[Page 66710]]
may be taken to ``Shutdown'' in order to scram the plant. The
response of the plant is the same as it is under the current plant
shutdown procedures. There are no other differences in testing
techniques or testing criteria from those previously required by the
NMP1 Technical Specifications. No physical change to the plant
design, materials, or standards is involved. No change to
instrumentation operating characteristics outside current tolerances
will be made. No unexpected plant transients will be initiated as a
result of the proposed changes. No initiator of any accident
previously evaluated is adversely affected. No system required to
actuate to respond to any accident previously evaluated in the UFSAR
is adversely affected by the proposed changes.
Category 4: The associated system or equipment will be isolated
or out of service during the performance of the required
surveillances. The associated surveillances will be performed during
quarterly valve stroking, at which time the system or equipment is
already out of service. No physical change to the plant design,
materials, or standards is involved. No change to instrumentation
operating characteristics outside current tolerances will be made.
No plant transients will be initiated as a result of the proposed
changes. No initiator of any accident previously evaluated is
adversely affected. No system required to actuate to respond to any
accident previously evaluated in the UFSAR is adversely affected by
the proposed changes.
The correction of the typographical errors is administrative only
and has no affect on plant systems or procedures. In all cases,
equipment used for accident mitigation is not adversely affected. The
ability of the operators to safely shut down NMP1 is not impaired. The
changes will not adversely affect any accident precursor or initiator
of any accident. For these reasons, the proposed changes will not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Each of the four categories [* * *] are evaluated separately below.
Category 1: The associated systems will remain operable and able
to automatically perform as designed any required safety functions
that may become necessary during performance of required
surveillances. No physical change to the plant design, materials, or
standards is involved. No change to instrumentation operating
characteristics outside current tolerances will be made. No accident
initiator or failure of a different type than previously identified
in the UFSAR is introduced. No different or new plant transients may
result from those previously evaluated in the UFSAR.
Category 2: The associated systems will be required for
monitoring purposes only and provide no automatic safety actuation
function. Since redundant, or redundant and alternate monitoring
instrumentation will still be available as required by the technical
specifications, the associated systems' functions in accident
mitigation are not affected. No physical change to the plant design,
materials, or standards is involved. No change to instrumentation
operating characteristics outside current tolerances will be made.
No accident initiator or failure of a different type than previously
identified in the UFSAR is introduced. No different or new plant
transients may result from those previously evaluated in the UFSAR.
Category 3: There will be no change in the system configuration
or plant operating conditions during the performance of associated
surveillances. The associated system's ability to perform required
safety functions will not be affected, whether the plant is shutdown
for refueling or shutdown for maintenance. The surveillances
performed to meet the requirements of NMP1 Technical Specifications
Tables 4.6.2a Parameter 8 and 4.6.2g Parameter 6 are included in
this category and may also be performed in concurrence with a unit
shutdown. The only difference between the proposed changes and the
normal unit shutdown sequence is that the mode switch may be taken
to ``Shutdown'' in order to scram the plant. The response of the
plant is the same as it is under the current plant shutdown
procedures. There are no other differences in testing techniques or
testing criteria from those previously required by the NMP1
Technical Specifications. No physical change to the plant design,
materials, or standards is involved. No change to instrumentation
operating characteristics outside current tolerances will be made.
No unexpected plant transients will be initiated as a result of the
proposed changes. No accident initiator or failure of a different
type than previously identified in the UFSAR is introduced. No
different or new plant transients may result from those previously
evaluated in the UFSAR.
Category 4: The associated system or equipment will be isolated
or out of service during the performance of the required
surveillance. The associated surveillances will be performed during
quarterly valve stroking, at which time the system or equipment is
already out of service. No physical change to the plant design,
materials, or standards is involved. No change to instrumentation
operating characteristics outside current tolerances will be made.
No plant transients will be initiated as a result of the proposed
changes. No accident initiator or failure of a different type than
previously identified in the UFSAR is introduced. No different or
new plant transients may result from those previously evaluated in
the UFSAR.
The correction of the typographical errors is administrative only
and has no affect on plant systems or procedures. In all cases, the
changes will not adversely affect any accident precursor or initiator
of any accident and, therefore, the changes do not introduce any new
failure modes or conditions that may create a new or different
accident. For these reasons, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated in the UFSAR.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
Each of the four categories [* * *] are evaluated separately below.
Category 1: The associated systems will remain operable and able
to automatically perform required safety functions during
performance of surveillances that satisfy the surveillance
requirement. There will be no effective change in the interval of
the affected surveillances. The probability of instrument drift or
the ability to detect a failed or drifted instrument remains
unchanged. No physical change to the plant design, materials, or
standards is involved. No change to instrumentation operating
characteristics outside current tolerances will be made. No system
required to actuate to respond to any accident is adversely affected
by the proposed changes. Since each system's operability is not
affected, the margin of safety associated with these systems will
not be significantly reduced.
Category 2: The associated systems will be required for
monitoring purposes only and provide no automatic safety actuation
function. Redundant, or redundant and alternate monitoring
instrumentation will still be available as required by the technical
specifications during the performance of the associated
surveillances. No physical change to the plant design, materials, or
standards is involved. No change to instrumentation operating
characteristics outside current tolerances will be made. There will
be no effective change in the intervals of the affected
surveillances. The probability of instrument drift or the ability to
detect a failed or drifted instrument remains unchanged. No plant
transients will be initiated as a result of the proposed changes. No
initiator of any accident previously evaluated is adversely
affected. No system required to actuate to respond to any accident
is adversely affected by the proposed changes. Therefore, the
associated systems' functions in accident mitigation are not
affected, and no margin of safety will be significantly reduced.
Category 3: There will be no change in the system configuration
or plant operating conditions during the performance of associated
surveillances, the associated system's ability to perform required
safety functions will not be affected, whether the plant is shutdown
for refueling or shutdown for maintenance. The surveillances
performed to meet the requirements of NMP1 Technical Specifications
Tables 4.6.2a Parameter 8 and 4.6.2g Parameter 6 may also be
completed in concurrence with a unit shutdown. The only difference
between the proposed changes and the normal unit shutdown sequence
is that the mode switch may be taken to ``Shutdown'' in order to
[[Page 66711]]
scram the plant. The response of the plant is the same as it is
under the current plant shutdown procedures. There are no other
differences in testing techniques or testing criteria from those
previously required by the NMP1 Technical Specifications. No
physical change to the plant design, materials, or standards is
involved. No change to instrumentation operating characteristics
outside current tolerances will be made. There will be no effective
change in the intervals of the affected surveillances. The
probability of instrument drift or the ability to detect a failed or
drifted instrument remains unchanged. No unexpected plant transients
will be initiated as a result of the proposed changes. No initiator
of any accident if adversely affected. No system required to actuate
to respond to any accident previously evaluated is adversely
affected by the proposed changes. Therefore, no margin of safety
will be significantly reduced.
Category 4: The associated system or equipment will be isolated
or out of service during the performance of the required
surveillances. The associated surveillances will be performed during
quarterly valve stroking, at which time the system or equipment will
already be out of service. No physical change to the plant design,
materials, or standards is involved. No change to instrumentation
operating characteristics outside current tolerances will be made.
There will be no effective change in the intervals of the affected
surveillances. The probability of instrument drift or the ability to
detect a failed or drifted instrument remains unchanged. No plant
transients will be initiated as a result of the proposed changes. No
accident initiator or failure of a different type than identified in
the UFSAR is introduced. Therefore, no margin of safety will be
significantly reduced.
The correction of the typographical errors is administrative only
and has no affect on plant systems or procedures. In all cases, the
changes will not adversely affect any accident precursor or initiator
of any accident and, therefore, the changes do not introduce any new
failure modes or conditions that may create a new or different
accident. None of the proposed changes involve physical modification of
the plant or alterations to any accident or transient analysis.
Therefore, for this and the above reasons, these proposed changes do
not involve any significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa, Acting Director.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: October 16, 1996.
Description of amendment request: The proposed amendment would
change certain requirements stated in Technical Specification 3/4.8.1,
``AC Sources''. The requirements are related to the emergency diesel
generators (EDGs). The proposed changes would:
1. Increase the EDG fuel storage system minimum volume requirements
specified in Limiting Condition for Operation 3.8.1.1.b.2;
2. Add a footnote applicable to Surveillance Requirement
4.8.1.1.2.f to qualify the words during shutdown. The footnote would
allow the option of performing selected surveillances, or portions
thereof, during conditions or modes other than shutdown;
3. Delete from Surveillance Requirement 4.8.1.1.2.f.14 the
requirement to verify that the cooling tower fans start automatically
on a Tower Actuation signal; and
4. Delete Surveillance Requirement 4.8.1.1.2.h.2 which specifies
performing a periodic pressure test on the ASME Code Class 3 diesel
fuel oil piping.
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)).
1. Limiting Condition for Operation 3.8.1.1.b.2
The proposed change increases the minimum EDG fuel oil storage
requirement to account for various factors that may affect the fuel
consumption rate. The revised storage requirement reflects actual EDG
test data and accounts for external variables including fuel oil
specific gravity, heating value of the fuel, and ambient conditions.
The proposed increase in the minimum volume storage requirement is
conservative and ensures that there will be at least a 7 day supply of
fuel oil stored for each EDG to meet the maximum Engineered Safety
Feature load requirements following a loss of power and a design basis
accident as described in Updated Final Safety Analysis Report (UFSAR)
Section 9.5.4.1, Diesel Generator Fuel Oil Storage and Transfer
System--Design Basis. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. Surveillance Requirement 4.8.1.1.2.f
The proposed change qualifies the requirement to perform EDG
surveillance requirements ``during shutdown''. Because the terms Hot
Shutdown and Cold Shutdown are defined in the TSs as operating modes or
conditions, the requirement to perform certain surveillances during
shutdown may be misinterpreted, as noted in NRC Generic Letter 91-04.
The proposed footnote would permit certain maintenance and testing
activities to be performed during conditions or modes other than
shutdown. The proposed footnote to Surveillance Requirement 4.8.1.1.2.f
would not alter the intent or the method by which the surveillances are
conducted, and the acceptance criteria for the surveillances would be
unchanged. The footnote would not degrade the ability of the EDGs to
perform their intended function, and it would not affect the response
of the EDGs to a loss of power as described in the UFSAR. Since plant
response to an accident would not change and since failure of an EDG
could not initiate any of the accidents evaluated in the UFSAR, the
proposed footnote would not alter the probability or consequences of an
accident previously analyzed.
3. Surveillance Requirement 4.8.1.1.2.f.14
The cooling tower functions as the ultimate heat sink following a
seismic event which results in blockage of the circulating water
tunnels and therefore a loss of service water. Amendment 18 eliminated
the requirement for automatic start of the cooling tower fans;
therefore, the automatic-start function for the cooling tower fans has
been defeated by placing the control switch in ``Pull-to-Lock''. The
proposed change to delete the automatic fan start reference from
Surveillance Requirement 4.8.1.1.2.f.14 is administrative only to
correct an oversight since the requirement should have been deleted
with the issuance of Amendment 18. The proposed deletion does not
affect the manner by which the facility is operated or involve any
[[Page 66712]]
changes to equipment or features which affect the operational
characteristics of the facility. Since there is no change to the
facility or operating procedures, there is no affect upon the
probability or consequences of any accident previously analyzed.
4. Surveillance Requirement 4.8.1.1.2.h.2
The ASME Code, Section XI, including applicable ASME Code Cases as
authorized by the NRC, provides alternate test methods to use in lieu
of a 110% hydrostatic pressure test that is not practical to perform on
the EDG fuel oil system as currently designed. With the proposed
deletion of Surveillance Requirement 4.8.1.1.2.h.2, the provisions of
Surveillance Requirement 4.0.5 and the ASME Code along with NRC-
authorized Code Cases would be utilized as an equivalent testing
requirement to ensure the continued integrity of the diesel fuel oil
system. Therefore, since the reliability of the EDG fuel oil system
will not be reduced, the probability or consequences of any accident
previously evaluated is not increased.
B. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)).
1. Limiting Condition for Operation 3.8.1.1.b.2
The proposed minimum fuel storage requirement has been developed
using actual EDG performance data and accounting for possible
variations in fuel oil specific gravity, heating value of the fuel, and
ambient conditions. The proposed change will provide additional
assurance that there will be at least a 7 day supply of fuel oil to
meet the maximum Engineered Safety Feature load requirements following
a loss of power and a design basis accident. The amount of fuel oil
stored has no effect upon the initiation of any accident sequence,
therefore, the proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed.
2. Surveillance Requirement 4.8.1.1.2.f
The proposed change to allow the option (as supported by a 10 CFR
50.59 safety evaluation) of performing selected surveillance tests, or
portions thereof, during conditions or modes other than during shutdown
does not affect the operation or response of any plant equipment,
including the EDGs, or introduce any new failure mechanism. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any previously analyzed.
3. Surveillance Requirement 4.8.1.1.2.f.14
Amendment 18 to the Seabrook Station Operating License approved the
change in the cooling tower operating mode from automatic actuation to
manual actuation. The proposed change to Surveillance Requirement
4.8.1.1.2.f.14 does not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because it does not affect the manner by which the
facility has been operated since Amendment 18 was issued, involve any
changes to equipment or features which affect the operational
characteristics of the facility, or introduce a new failure mode. The
proposed change merely corrects an oversight in that the requirement
should have been deleted when Amendment 18 was issued.
4. Surveillance Requirement 4.8.1.1.2.h.2
The change does not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because it does not affect the manner by which the
facility is operated as assumed in the design analysis or Safety
Evaluation, involve any changes to equipment or features which affect
the operational characteristics of the facility, or introduce a new
failure mode. The proposed change merely provides a practical alternate
test method using methods acceptable per Section XI of the ASME Code,
applicable ASME Code Cases as authorized by the NRC, and Regulatory
Guide (RG) 1.137, ``Fuel-Oil Systems at Nuclear Power Plants,''
Revision 1, October 1979. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from any
previously analyzed.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)).
1. Limiting Condition for Operation 3.8.1.1.b.2
The proposed change does not reduce the ability of the EDGs to
provide sufficient power for at least 7 days to meet the maximum
Engineered Safety Feature load requirements following a loss of power
and a design basis accident as described in UFSAR Section 9.5.4.1.
2. Surveillance Requirement 4.8.1.1.2.f
The proposed change does not reduce the ability of the EDGs to
provide sufficient power to meet the maximum Engineered Safety Feature
load requirements following a loss of power and a design basis accident
as described in the UFSAR. Performing certain surveillances during
conditions or modes other than shutdown (as supported by a 10 CFR 50.59
safety evaluation) does not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because it does not affect the manner by
which the facility is operated as assumed in the design analysis or
Safety Evaluation, involve any changes to equipment or features which
affect the operational characteristics of the facility. The proposed
change will continue to ensure the reliability of the EDGs to perform
their intended function.
3. Surveillance Requirement 4.8.1.1.2.f.14
The change does not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because it does not affect the manner by which the
facility has operated since Amendment 18 was issued, involve any
changes to equipment or features which affect the operational
characteristics of the facility, or introduce a new failure mode. The
proposed change merely corrects an oversight in that the requirement
should have been deleted when Amendment 18 was issued.
4. Surveillance Requirement 4.8.1.1.2.h.2
The change does not involve a significant reduction in a margin of
safety (10 CFR 50.92(c)(3)) because it does not affect the manner by
which the facility is operated or involve any changes to equipment or
features which affect the operational characteristics of the facility.
The proposed change will continue to ensure the reliability of the EDG
fuel oil system. The proposed change merely provides a practical
alternate test method using methods acceptable per Section XI of the
ASME Code, applicable ASME Code Cases as authorized by the NRC, and
Regulatory Guide (RG) 1.137, ``Fuel-Oil Systems at Nuclear Power
Plants,'' Revision 1, October 1979.
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
[[Page 66713]]
Service Company, Post Office Box 270, Hartford CT 06141-0270.
NRC Project Director: S. Singh Bajwa, Acting.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: October 17, 1996.
Description of amendment request: The proposed amendment would
delete certain instrumentation requirements stated in Technical
Specification (TS) 3/4.3, Instrumentation. The deleted requirements
would be relocated to the Seabrook Station Technical Requirements
Manual (SSTR). The associated Bases for the deleted TS requirements
will be deleted also, but they will not be incorporated into the SSTR.
The following Limiting Conditions for Operation (LCO) and associated
Surveillance Requirements (SRs) would be relocated to the SSTR:
------------------------------------------------------------------------
Technical specification Title
------------------------------------------------------------------------
LCO--3.3.3.2.............................. Incore Detector System.
LCO--3.3.3.3 and associated SRs & Tables.. Seismic Instrumentation.
LCO--3.3.3.4 and associated SRs & Tables.. Meteorological
Instrumentation
LCO--3.3.4 and associated SRs............. Turbine Overspeed
Protection.
------------------------------------------------------------------------
The proposed amendment would also delete (without relocating to the
SSTR) the reference to the location of the meteorological tower from TS
5.5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)) because the proposed changes do not involve any physical
changes to the plant, do not alter the way any structure, system or
component functions, do not modify the manner in which the plant is
operated, do not impact the physical protective boundaries of the
plant, and do not decrease the effectiveness of administrative controls
for assuring safe operation of the facility. The instrumentation-
related systems are not considered a design feature or an operating
restriction that is an initial condition of a design basis accident or
transient analysis, nor do they function in any way to mitigate the
consequences of a design basis accident or transient.
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 the proposed changes do not involve any physical
changes to the plant, do not alter the way any structure, system or
component functions, do not modify the manner in which the plant is
operated, do not impact the physical protective boundaries of the
plant, and do not decrease the effectiveness of administrative controls
for assuring safe operation of the facility.
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
involve any physical changes to the plant, do not alter the way any
structure, system or component functions, do not modify the manner in
which the plant is operated, do not impact the physical protective
boundaries of the plant, and do not decrease the effectiveness of
administrative controls for assuring safe operation of the facility.
Further, the proposed changes do not affect the ability of systems,
structures or components important to safety to perform their intended
function.
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: S. Singh Bajwa, Acting.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: March 20, 1996 and as supplemented on
July 25, 1996.
Description of amendment request: The amendments would modify the
Susquehanna Steam Electric Station (SSES), Units 1 and 2, Technical
Specifications to change the ``open'' logic for the high pressure core
injection (HPCI) suction valves HV-155/255-F042 in order to eliminate
the HPCI pump auto-transfer on high suppression pool level.
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.
Based on the following discussion for the containment, reactor
building, HPCI and RCIC [reactor core isolation cooling] systems,
and the safety-related valves in piping connected to the suppression
pool, the proposed action does not increase the probability or
consequences of an accident previously evaluated. Primary
Containment and Reactor Building Safety-Related Systems, Structures,
and Components Affected by LOCA/SRV [Loss-of-coolant-accident/safety
relief valve] Hydrodynamic Loads
As discussed in the Safety Assessment for this change,
elimination of the HPCI auto suction transfer on high suppression
pool level will allow higher suppression pool water levels in
accidents and transients which involve HPCI operation. The impact of
the higher suppression pool levels were examined for the following
design-basis accidents and transients:
Loss of Coolant Accidents inside containment (FSAR [Final Safety
Analysis Report] *6.2.1.1.3.3),
Inadvertent Safety/Relief valve opening (FSAR *15.1.4),
Primary system break outside containment (FSAR *3.6A),
Inadvertent HPCI initiation (FSAR *15.5.1),
Loss of feedwater flow (FSAR *15.2.7),
Loss of Offsite AC Power (FSAR *15.2.6),
Loss of Main Condenser vacuum (FSAR *15.2.5),
Inadvertent MSIV closure (FSAR *15.2.4),
Turbine trip (with and without bypass) (FSAR *15.2.3),
Generator Load Rejection (with and without bypass), (FSAR
*15.2.2), and
Pressure regulator failure-closed/open (FSAR *15.2.1 & 15.1.3).
These accidents and transients were selected for evaluation
because they involve an initiation of the HPCI system either
inadvertently or as a result of a decrease in vessel inventory and/
or coolant level. Two special events, ATWS and SBO, are also
considered along with the design basis events listed above.
It was concluded that design-basis SRV and LOCA loads envelop
the loads expected with the proposed change. Therefore, the proposed
change does not increase the failure probability of any primary
containment or reactor building structure, system or component which
is affected by LOCA/SRV hydrodynamic loads. The major findings which
lead to this conclusion about SRV and LOCA loads are summarized
below:
DBA [design basis accident] dynamic pressure loads are based on
a maximum initial suppression pool level of 24 feet. The proposed
modification to the HPCI suction
[[Page 66714]]
transfer logic does not affect the initial pool level or the initial
suppression chamber air space volume. During normal plant operation,
suppression pool level (and hence suppression chamber air space
volume) is controlled by Technical Specification requirements.
For LOCAs other than the DBA, the containment is designed for
ADS [automatic depressurization system] blowdown loads in
combination with the LOCA loads. For an intermediate break, the
proposed HPCI modification does allow suppression pool level to
exceed 24 feet by a small amount. ADS loads are, however,
independent of suppression pool level when the downcomer vents are
cleared. Therefore, the proposed modification has no influence on
ADS hydrodynamic loads for an intermediate break.
For small breaks, HPCI injection prevents ADS actuation.
Nevertheless, SRV actuations occur during the RPV [reactor pressure
vessel] cooldown. Downcomer vents are opened in the beginning part
of the accident, but close later on as the break enthalpy decreases.
When the downcomer vents are cleared, the level inside the SRV
tailpipe is not influenced by pool level, and therefore, the SRV
hydrodynamic loads are unaffected by the proposed modification.
During the phase of the accident in which the downcomer vents are
sealed with water, there are no wetwell LOCA hydrodynamic loads, but
the SRV loads are dependent on SP [suppression pool] water level. In
this case, SRV loads are acceptable because SP water level is always
below the Load Limit curve.
ADS actuation would be required in the event of a HPCI failure
during a small-break accident. If HPCI fails during the phase of the
accident in which the downcomer vents are cleared, then ADS loads
would be acceptable because water level (and air volume) within the
SRV tailpipes is independent of pool level. Even if HPCI failure
occurs in the latter part of the accident where the downcomer vents
are sealed, ADS loads are acceptable because water level is always
well below the Load Limit curve.
Under non-LOCA conditions, the containment is designed for
simultaneous actuation of all 16 SRVs. The Load Limit Line defines
the acceptable operating region, in terms of reactor pressure and
suppression pool level, for actuation of all 16 SRVs. Following a
plant transient involving HPCI operation, the suppression pool level
is always below the Load Limit curve, and only a small number of
SRVs actuate to remove decay heat from the reactor.
HPCI System
The proposed change does not increase the probability of an
equipment malfunction in the HPCI system. In fact, the change
eliminates the potential failure of the HPCI suction auto-transfer
on high suppression pool level since that logic is removed.
Potential spurious auto-transfer associated with high suppression
pool logic is also eliminated. HPCI suction auto-transfer on low CST
[condensate storage tank] level and its potential to fail are
unchanged by this change. Also, the change does not affect the
manual suction transfer from the CST to the suppression pool.
As discussed in the safety assessment for this change, the
proposed change has no adverse effects on HPCI valves, pump, or
turbine. Therefore, elimination of the HPCI suction auto transfer
logic (on high suppression pool level) does not increase the
probability of a HPCI malfunction. The consequence of a HPCI failure
in a design-basis accident is evaluated in NEDC-32071P Rev.1,
``Susquehanna Steam Electric Station Units 1 and 2 SAFER/GESTR-LOCA
Loss-of-Coolant Accident Analysis.'' With regard to the fuel, the
consequence of a HPCI failure is unaffected by the proposed change.
If HPCI fails in a design-basis small break accident, ADS
actuation would be required. ADS loads continue to be enveloped by
design loads with the proposed change. Therefore, the proposed
change does not increase the consequences of a HPCI failure.
HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center
1D264(2D264)
On a component level, the failure probability and consequences
of failure associated with the AX [auxiliary] relay in 250 VDC
Control Center 1D264 (2D264) are eliminated because the relay is
disconnected and removed by this modification. Since the control
functions of K19 in panel 1C620 (2C620) have been eliminated, the
failure of the relay has no effect on HPCI suction valve F042
operation.
The 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel
1C620 (2C620) both receive power from battery systems during Station
Blackout. Removal of the relay from 250 VDC Control Center 1D264
(2D264) and the replacement of the relay in HPCI Relay Panel 1C620
(2C620) decreases the load on the battery systems by a small amount.
The change in battery load and line voltage drop is negligible and
is documented in applicable calculations. Dynamic qualification of
the subject equipment is not adversely affected by this modification
as documented in applicable calculations.
RCIC Turbine
As discussed in the safety assessment for this change, RCIC is
used to provide coolant makeup following a reactor vessel isolation
and for an Appendix R shutdown scenario. The Appendix R event also
assumes the reactor vessel is isolated. These events are discussed
in Section 15.2.4 of the FSAR and in the FPRR [fire protection
review report]. The proposed change has no adverse effects on RCIC
turbine operation following a MSIV [main steam isolation valve]
closure (see discussion in the safety assessment for this change
[letter dated March 20, 1996, as supplemented July 25, 1996]).
Therefore, there is no increase in the RCIC failure probability for
the MSIV-closure event or the Appendix R shutdown scenario. The
consequence of RCIC failure is unchanged by the proposed
modification; if RCIC fails, HPCI is available as a backup
system.\1\ [All footnotes are listed at the end of the no signficant
hazards basis section.]
Although RCIC is not designed for mitigation of a small break
accident, the effect of the proposed change on RCIC turbine
operation for such an accident was evaluated in the safety
assessment for this change. The assessment concludes that the
proposed change has no adverse effects on RCIC operation, and
therefore, there is no increase in RCIC failure probability during a
small break accident. Failure of RCIC in a small break accident
would require ADS initiation only for a particular break flow which
is slightly greater than HPCI injection capability. But ADS
initiation has already been considered when evaluating the
consequences of HPCI failure during a small break accident.
Safety-Related Valves on Piping Connected to Suppression Chamber
MOVs [motor operated valves]--The proposed change could
potentially lead to a maximum suppression pool level of 26 feet in a
design-basis accident. This is 2 feet above the maximum design level
of 24 feet. As discussed in the safety assessment for this change,
this is equivalent to a pressure increase of 0.86 psi at the bottom
of the suppression pool. This small pressure increase has negligible
effect on valve operation, and therefore, there is no increase in
the probability of a failure or malfunction of valves in piping
connected to the suppression pool.
Vacuum Breakers--Allowing suppression pool level to potentially
increase to 26 feet in a design-basis accident does not affect the
failure probability of downcomer-vent vacuum breakers because the
level is well below the vacuum breaker elevation of 42 feet.
SRVs/Tailpipes--As discussed in the safety assessment for this
change, the increased suppression pool level associated with the
proposed change does not have any adverse effect on SRV operation or
on the structural integrity of the SRV tailpipe.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Based on the following discussion for the containment, reactor
building, HPCI and RCIC systems, and the safety-related valves in
piping connected to the suppression pool, the proposed action does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The following discussion concerning the impact of the change on
the primary containment, the reactor building, the HPCI system, and
safety-related valves, provides the basis for this conclusion.
Primary Containment and Reactor Building Safety-Related Systems,
Structures, and Components Affected by LOCA/SRV Hydrodynamic Loads
The HPCI suction transfer logic is not necessary to maintain
LOCA loads within design limits because these dynamic pressure loads
are characterized in terms of the SP level at the initiation of the
accident. That is, LOCA blowdown tests were conducted without the
removal of water from the suppression chamber section of the test
tank.2 The increase in pool level realized during these tests
was proto-typical of the pool level increase expected at
Susquehanna. Removal of the HPCI suction transfer logic on high pool
level does not affect suppression pool level at the initiation of a
DBA.3
[[Page 66715]]
In addition, the HPCI suction transfer logic is not necessary to
maintain SRV/ADS blowdown loads within design limits. SRV dynamic
pressure loads consist of two components: air clearing loads and
steam condensation loads. The steam condensation loads are bounded
by the more severe air clearing loads which are caused by gas bubble
oscillations following the expulsion of noncondensible gas from the
SRV tailpipe. Air clearing loads are a function of reactor pressure
and water level inside the SRV tailpipe.
Depending on the break size and location, the downcomer vents
may be cleared for the entire time that HPCI is operating, or they
may reseal in the latter part of the accident. When the downcomer
vents are cleared, the level inside the SRV tailpipe is depressed to
the elevation coinciding with the bottom of the downcomer pipes, and
it is therefore decoupled from the rising suppression pool level. In
this situation SRV air-clearing loads are unaffected by the proposed
change.
When the downcomer vents are sealed with water, the Load Limit
line can be used to determine if SRV/ADS loads are enveloped by
design loads. For the most limiting event, which is the small break
LOCA, the overall safety margin increases as pool level rises during
the event. This is because the decrease in reactor pressure more
than offsets the adverse effects associated with the rise in pool
level.
Since LOCA and SRV dynamic loads remain bounded by design loads,
dynamic loading of primary containment and reactor building
structures, systems, and components are unaffected by the proposed
change. Therefore, with respect to dynamic loads, the proposed
change does not create the possibility for an accident or
malfunction of a different type than any evaluated in the SAR
[Safety Analysis Report].
HPCI System
There are no new HPCI turbine failure modes introduced by the
higher suppression pool levels which can occur with the proposed
change. Turbine exhaust pressure remains well below the design limit
of 65 psia. In addition, the higher pool level does not create the
possibility of water hammer damage to the turbine discharge piping.
If the operator fails to control RPV level less than +54'' (single
operator error) in the long-term part of the small-break accident
when suppression pool level is greater than 25.6 feet, leakage
through check valve F049 is such that it will be contained well
within the volume of the turbine-discharge-line drain pot. Note that
suppression pool level is limited to 26 feet by operator action.
Furthermore, suppression pool level can reach 26 feet only for a
particular range of small breaks, and for this range of small
breaks, suppression pool level would exceed 25.6 feet for only
approximately 10 minutes of the accident duration. This corresponds
to about 10% of the time that HPCI is operating. Thus it is very
unlikely that HPCI would trip with pool level greater than 25.6
feet.
If check valve F049 is failed during the small-break accident
(single equipment failure), the turbine exhaust line would become
flooded if the HPCI system tripped during the 10 minute interval
when suppression pool level greater than 26 feet; however, it is not
necessary to postulate an operator error (failure to control RPV
level less than +54'') along with the check valve failure. A small
break accident with failure of check valve F049 and failure of the
operator to control RPV level as required by the EOPs [emergency
operating procedures], in a narrow time interval during the long-
term part of the accident, is beyond the plant design basis.
A new type of malfunction does not occur even in the beyond-
design-basis condition where failure of check valve F049 is
considered along with failure of the operator to control RPV level
less than 54'' in the narrow time interval when pool level is
greater than 25.6. With these failures, the turbine exhaust piping
will become flooded, and the system may fail on restart. The General
Electric Company has performed an analysis to determine the
consequences of a HPCI start with flooding of the turbine and
adjacent exhaust line.\4\ The analysis, which addresses a potential
design deficiency in the HPCI barometric condenser, shows that the
containment penetration head fitting and interface piping will not
fail as a result of the water hammer associated with the HPCI start.
Since failure of the HPCI system is already considered in the plant
design-basis accident analysis; this is not a different type of
malfunction than that already considered.
HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center 1D264(2D264)
No new failure modes are introduced by the hardware changes in
the 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel 1C620
(2C620). Some failure modes are eliminated by the proposed change.
Specifically, the potential failure of the HPCI suction auto-
transfer on high suppression pool level is eliminated since that
logic is removed. Potential spurious auto-transfer associated with
high suppression pool logic is also eliminated. HPCI suction auto-
transfer on low CST level and its potential to fail are unchanged by
this change.
On a component level, potential failure modes for the AX relay
in 250 VDC Control Center 1D264 (2D264) are eliminated by this
modification because the relay is disconnected and removed by this
change. The potential failure modes for the relay K19 in panel 1C620
(2C620) are unchanged. Since the control functions of K19 have been
eliminated, the failure of the relay has no effect on HPCI suction
valve F042 operation.
Removal of the relay from 250 VDC Control Center 1D264 (2D264)
and the replacement of the relay in the HPCI Relay Panel 1C620
(2C620) changes the load on the battery systems by a small amount.
The change in battery load and change in line voltage drop are
negligible and they do not adversely affect the performance of the
panels or battery systems. In addition, seismic qualification of the
panels is not adversely affected by this change.
RCIC Turbine
As discussed in the safety assessment for this change, the
proposed change has no adverse effects on RCIC turbine operation.
Therefore, the proposed change cannot result in a new RCIC failure
mode.
Safety-Related Valves on Piping Connected to Suppression Chamber
MOVs--The increased suppression pool water level which can occur
as a result of the proposed change does not create a failure
mechanism for safety-related valves on piping connected to the
suppression pool. The pressure differential for any valve on piping
connected to the suppression pool will increase by at most 0.86 psi.
This change in differential pressure has negligible effect on valve
operation.
Vacuum Breakers--The proposed change cannot lead to malfunction
of the downcomer-vent vacuum breakers as the maximum level expected
in a design-basis event is 26 feet, and the vacuum breakers are
located at 42 feet above the suppression pool floor.
SRVs/Tailpipes--There is no interaction between increased
suppression pool level and SRV operation since the flow through the
SRVs is choked and therefore decoupled from downstream conditions.
Also, the increased suppression pool level cannot lead to failure of
the SRV tailpipe because the potential level increase is well below
the SRV Tailpipe Level Limit.\5\ If suppression pool water level is
below this limit, there is no concern of tailpipe failure due to
overpressurization. The minimum value of the SRV Tailpipe Level
Limit is 35 feet.\6\ This is 9 feet above the maximum level expected
in a design-basis accident. For beyond-design-basis events, SRV
tailpipe integrity is protected by the EOP requirement to
depressurize the reactor on the SRV Tailpipe Level Limit.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Based on the following discussion for the containment, reactor
building, HPCI and RCIC system, and the safety-related valves in
piping connected to the suppression pool, the proposed action does
not involve a significant reduction in a margin of safety.
HPCI System
The HPCI Technical Specifications ensure that the system is
capable of providing adequate core cooling to limit clad
temperatures in the event of a small break LOCA which does not
result in rapid depressurization of the RPV (Technical Specification
Section 3/4.5.1 & 3/4.5.2). The proposed change has no adverse
affects on the injection capability of the HPCI system. Therefore,
the safety function of the system is not degraded, and there is no
reduction in the margin of safety as defined in the basis for the
HPCI Technical Specifications.
Primary Containment and Reactor Building Safety-Related Systems,
Structures, and Components Affected by LOCA/SRV Hydrodynamic Loads
Removal of the HPCI auto suction transfer on high suppression
pool level does not affect the Technical Specification requirement
to maintain suppression pool water level between 22 and 24 feet
(Technical Specification 3.6.2.1). Therefore, the maximum
containment pressure during the design-basis accident is unaffected
by the proposed change, and there can be no reduction in the margin
of safety as defined in the basis for Technical Specification
[[Page 66716]]
3.6.2.1. Furthermore, a detailed examination of the reactor and
containment response under accident and transient conditions
involving HPCI operation found no situations where the auto suction
transfer was necessary to maintain LOCA and SRV loads within the
design basis envelope. Therefore, from the standpoint of LOCA/SRV
hydrodynamic loads, the proposed change does not reduce the margin
of safety for any primary containment or reactor building structure,
system, or component.
RCIC Turbine
The basis for Technical Specification 3.7.3 states that the RCIC
system is provided to assure adequate core cooling in the event of a
reactor isolation with loss of feedwater flow. The proposed change
does not prohibit RCIC from performing this function, nor does it
degrade in any way the core cooling capability of RCIC. Therefore,
there is no reduction in the margin of safety as defined in the
basis for Technical Specification 3.7.3.
Safety-Related Valves on Piping Connected to Suppression Pool
MOVs--The increase in suppression pool water level which can
occur as a result of the proposed change does not reduce the margin
of safety for safety-related valves on piping connected to the
suppression pool. The pressure differential for any valve on piping
connected to the suppression pool will increase by at most 0.86 psi.
This change in differential pressure has negligible effect on valve
operation.
Vacuum Breakers--The proposed change cannot reduce the margin of
safety as discussed in the basis for Technical Specification 3.6.4
because the maximum level expected in a design-basis event is 26
feet which is well below the downcomer-vent vacuum breaker elevation
of 42 feet.
SRVs/Tailpipes--There is no interaction between increased
suppression pool level and SRV operation since the flow through the
SRVs is choked and therefore decoupled from downstream conditions.
Consequently, there is no reduction in the margin of safety as
defined in the bases for Technical Specifications 3.4.2 (safety
valve function) and 3.5.1.d (ADS function). Also, the increased
suppression pool level does not lead to a reduction in the margin of
safety for the SRV tailpipes because the tailpipes can operate
safely with pool levels up to 35 feet. This is nine feet above the
maximum suppression pool level that can occur in a design-basis
accident with the proposed change. For beyond-design-basis events,
SRV tailpipe integrity is protected by the EOP requirement to
depressurize the reactor on the SRV Tailpipe Level Limit.\7\
HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center 1D264(2D264)
As discussed previously, removal of the relay from 250 VDC
Control Center 1D264 (2D264) and the replacement of the relay in the
HPCI Relay Panel 1C620 (2C620) changes the load on the battery
systems by a small amount. The change in battery load and change in
line voltage drop are negligible and therefore they do not reduce
the margin of safety for the panels or battery systems. In addition,
seismic qualification of the panels is not adversely affected by
this change so there is no reduction in the margin of safety for
seismic events.
1. DBD041, Rev. 0, p. 1. [design basis document for RCIC system]
2. SSES DAR [design assessment report for suppression pool
hydrodynamic loads], Section 9.4.1
3. Suppression pool level must be maintained less than 24 feet
in accordance with Technical Specification 3.6.2.1.a.
4. GKR-03-001, ``NRC and Utility Notification of Closeout of GE
PRC92-05, Potential Design Deficiency on HPCI,'' January 6, 1993 [GE
letter to PP&L regarding closure of HPCI design issue].
5. This limit is defined in EO-100/200-103 [emergency operating
procedure]
6. Bechtel Calculations PUP-15598-S2 & PUP-15598-S6, and PLE-
15315 (March 2, 1992)
7. The limit is defined in EO-100/200-103 [emergency operating
procedure]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: October 7, 1996.
Description of amendment request: The amendments would modify the
Susquehanna Steam Electric Station, Units 1 and 2, Technical
Specifications by revising the trip setpoints and allowable values for
the secondary containment isolation ``Refuel Floor High Exhaust Duct
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct
Radiation--High'' monitor in Table 3.3.2-2. The change would enhance
the operational efficiency of plant operations by eliminating
compensatory measures which prevent spurious secondary containment
isolations, and initiation of the standby gas treatment system (SGTS)
and recirculation system during refueling activities. This change would
also allow for the use of the hydrogen water chemistry system during
operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to the trip setpoints and allowable values
to the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,
the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor,
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor
does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The design basis
for the monitors is to monitor radiation in the unfiltered air from
the Zone III exhaust system to provide signals which isolate the
Zone III of the secondary containment on a high radiation condition,
and to initiate SGTS and the Recirculation system to limit offsite
doses to maintain regulatory requirements.
The original setpoints for these monitors were based upon normal
radiological operating conditions and were set at a value to
preclude spurious design actuations by these monitors during normal
plant operations. However, the monitors are designed to detect
radiation associated with certain postulated accident conditions. As
required by the Technical specifications the monitors are operable
when conditions exist that may result in fuel damage events, and
therefore, will perform their design basis function. Consequently,
an increase to the trip setpoints and allowable values is warranted
since the existing setpoints, which are conservatively based on
normal radiological operating conditions, are not related to the
design basis of the monitors. Therefore, based upon the design basis
of the monitors, an increase to the trip setpoints and allowable
values will not result in a decrease of the safety function of the
monitors but will make the trip setpoints and allowable values
consistent with the design basis.
Based on the design basis of these monitors, revised analytical
limits were derived reflecting the accident function of the
monitors. The analytical limit calculations utilized FSAR realistic
source terms, instead of the worst case source terms utilized for
10CFR [Part] 100 compliance. Use of the realistic source terms
results in conservative analytical limits.
The ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' [monitor]
are required to be OPERABLE during CORE ALTERATIONS (except for
single control rod movements unless performing TS 3.10.3),
operations with the potential for draining the reactor vessel, and
handling of irradiated fuel in the secondary containment. The
``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor is
required to be operable during handling of irradiated fuel. These
Technical Specification
[[Page 66717]]
applicable operational conditions for the monitors are not affected
since this proposed revision only revises the trip setpoints and
allowable values to be consistent with the design bases of the
monitors.
For the reasons stated above the revisions to the trip setpoints
and allowable values to the ``Refuel Floor High Exhaust Duct
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct
Radiation--High'' monitor in Technical Specification.
Table 3.3.2-2 can be implemented without a significant increase
in the probability or consequence of an accident previously
evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident previously evaluated.
The proposed change to the trip setpoints and allowable values
for the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,
the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor,
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The monitors are designed to limit the release of airborne
radioactivity in the secondary containment Zone III exhaust system
by isolating Zone III, initiating [the] SGTS and initiating the
Recirculation System on high radiation resulting from fuel handling
accidents. Therefore, the design basis for these monitors is to
monitor radiation in the unfiltered air from the Zone III exhaust
system, and provide signals to limit offsite doses to maintain
regulatory requirements. Zone III includes the Refueling Floor and
can include the Railroad Access Shaft during certain alignments.
These radiation monitors are not provided for occupational
protection associated with operational radiation doses. The proposed
revision does not affect the design basis of the monitors nor the
kind of accident associated with the basis; therefore, no potential
to create a new or different accident exists.
For the reasons stated above the revisions to the trip setpoints
and allowable values to the ``Refuel Floor High Exhaust Duct
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct
Radiation--High'' monitor in Technical Specification Table 3.3.2-2
can be implemented without creating the possibility of a new or
different kind of accident from any accident previously evaluated.
III. This proposal does not involve a significant reduction on a
margin of safety.
The proposed change to the trip setpoints and allowable values
for the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor,
the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor,
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor
does not involve a significant reduction in a margin of safety.
The monitors are designed to limit the release of airborne
radioactivity in the secondary containment Zone III exhaust system
by isolating Zone III, initiating [the] SGTS and initiating the
Recirculation System on high radiation resulting from fuel handling
accidents. Therefore, the design basis for these monitors is to
monitor radiation in the unfiltered air from the Zone III exhaust
system, and provide signals to limit offsite doses to maintain
regulatory requirements. Zone III includes the Refueling Floor and
can include the Railroad Access Shaft during certain alignments.
These radiation monitors are not provided for occupational
protection associated with operational radiation doses. However, the
original setpoints for these monitors were conservatively based upon
normal radiological operating conditions and were set at a value to
preclude spurious design actuation by these monitors during normal
plant operations. The calculations performed to support the trip
setpoint and allowable value revisions concluded that the change
will maintain offsite doses within the 10CFR100 limits. The ``Refuel
Floor High Exhaust Duct Radiation--High'' monitor, and the ``Refuel
Floor Wall Exhaust Duct Radiation--High'' are required to be
OPERABLE during CORE ALTERATIONS (except for single control rod
movements unless performing TS 3.10.3), operations with the
potential for draining the reactor vessel, and handling of
irradiated fuel in the secondary containment. The ``Railroad Access
Shaft Exhaust Duct Radiation--High'' monitor is required to be
operable during handling of irradiated fuel. These Technical
Specification applicable operational conditions for the monitors are
not affected since the proposed revision only revises the trip
setpoints and allowable values to be consistent with the design
bases of the monitors.
The proposed revisions to the trip setpoints and allowable
values, in addition to being based on the appropriate accident
conditions, were also developed utilizing standard setpoint change
methodologies that consider instrument and calibration accuracies
and instrument drift tolerances. This provides added conservatism to
assure that the revised trip setpoints and allowable values are not
exceeded.
For the reasons stated above the revisions to the trip setpoints
and allowable values to the ``Refuel Floor High Exhaust Duct
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct
Radiation--High'' monitor in Technical Specification Table 3.3.2-2
can be implemented without involving 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania.
Date of amendment request: November 25, 1996.
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise the wording in TS Section
4.8.1.1.2.e.2 and the associated TS Bases Section 3/4.8 to remove the
specific reference to the Residual Heat Removal pump motor and its
corresponding kW rating value, and replace it with wording consistent
with that specified in the Improved TS (i.e., NUREG-1433, Revision 1,
``Standard Technical Specifications General Electric Plants,'' dated
April 1995).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes do not make any physical alterations or
modifications to the plant systems or equipment. The proposed
changes do not adversely impact the operation of any plant
equipment. The EDGs will continue to function as designed to ensure
that the necessary electrical power is provided to essential plant
equipment to mitigate the consequences of an accident, e.g., Loss-
of-Offsite-Power (LOOP) and Loss-of-Coolant Accident LOCA)
coincident with a LOOP (LOCA/LOOP). The proposed TS changes do not
impact the performance testing requirements associated with the
EDGs. The accident mitigating capabilities of the diesel generators
and emergency loads will remain the same.
The proposed TS changes are consistent with the guidance
stipulated in NUREG-1433, Revision [1], ``Standard Technical
Specification General Electric Plants,'' regarding single load
rejection testing of the EDGs. Specifically, the proposed changes
involve revising the wording in TS Surveillance Requirement (SR)
4.8.1.1.2.e.2 to remove the specific reference to the Residual Heat
Removal (RHR) pump motor and associated kW loading value (992 kW),
and replace it with wording indicating that the EDGs must be capable
of rejecting the single largest post-accident load, which is
consistent with NUREG-1433, Revision 1, guidance. The proposed
changes will also provide additional flexibility for future plant
maintenance activities.
Each EDG will continue to be tested by rejecting a load of
greater than or equal to that of its single largest post-accident
load while maintaining voltage and frequency
[[Page 66718]]
within the current specified parameters. The RHR pump motors are
currently used in performing the EDG single load rejection testing.
The RHR pump motors will continued [sic] [continue] to be used in
performing the surveillance testing since they are the single
largest post-accident electrical load. The consequences of a
malfunction of equipment are not affected. Failure of a EDG or its
safety-related loads is bounded by the loss of a Class 1E electrical
power division which has been previously evaluated as discussed in
LGS Updated Final Safety Analysis Report (UFSAR) Sections 8.1.5.2.e
and 8.3.1.1.3.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes do not make any physical alterations or
modifications to the plant systems or equipment. The proposed
changes do not adversely impact the operation of any plant
equipment. The EDGs will continue to function as designed to provide
essential electrical power to mitigate the consequences of an
accident. The proposed TS changes are consistent with the guidance
stipulated in NUREG-1433, Revision 1, regarding single load
rejection testing of the EDGs. The proposed changes do not introduce
any new accidents or transients. The proposed TS changes will
provide additional flexibility for future maintenance activities.
The proposed changes do not alter any EDG testing requirements or
frequencies. The RHR pump motors are currently used in performing
the EDG single load rejection testing. The RHR pump motors will
continue to be used in performing the surveillance testing since
they are the single largest post-accident electrical load. The
operation of the EDGs and their corresponding safety-related
electrical loads remain unchanged as a result of the proposed TS
changes.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes do not involve any physical changes to
plant systems or equipment. The proposed TS changes are consistent
with the guidance stipulated in NUREG-1433, Revision 1, ``Standard
Technical Specification General Electric Plants,'' regarding single
load rejection testing of the EDGs. The proposed TS changes will
provide additional flexibility for future plant maintenance
activities. The EDGs will continue to function as designed to
provide essential electrical power to mitigate the consequences of
an accident. The operation of the EDGs and their corresponding
safety-related electrical loads remain unchanged as a result of the
proposed TS changes.
Therefore, the proposed TS changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101.
NRC Project Director: John F. Stolz.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: November 15, 1996.
Description of amendments request: The amendments would eliminate
the containment systems Technical Specification 3.6.2.2. ``Spray
Additive System.'' The specification would be replaced with a new
emergency core cooling system Technical Specification 3.5.6 ``ECCS
Recirculation Fluid pH Control System.''
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 proposed change involves replacement of concentrated
NaOH injected via the containment spray system with trisodium
phosphate (TSP) stored in the containment and dissolved in the sump
recirculation solution to maintain acceptable post accident spray/
recirculation solution chemistry. Deletion of the concentrated NaOH
will eliminate a personnel hazard. The pH control system functions
in response to an accident and does not involve or have any effect
on any initiating event for any accident previously evaluated.
Operation under the proposed amendments will continue to ensure that
iodine potentially released post-LOCA [loss-of-coolant accident] is
retained in the sump solution, and resultant offsite and control
room thyroid doses are within the limits of 10 CFR [Part] 100 and 10
CFR [Part] 50, Appendix A, General Design Criterion [GDC] 19,
respectively.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The deleted equipment is isolated from the remaining
equipment by cut-and-capped piping, determinated and/or spared
cables; and interfaces are analyzed to ensure the remaining required
equipment meets applicable original design requirements. The new
equipment (TSP and baskets) is a passive pH control system and is
supported and analyzed to ensure there are no adverse interfaces
(e.g., pipe break, jet impingement, seismic) with existing
equipment, system, or structures.
3. The proposed change does not involve a significant reduction
in a margin of safety. The slight change in recirculation solution
pH maintains adequate protection against chloride and caustic
induced stress corrosion cracking on mechanical systems and
components, and maintains the capability of the solution to retain
iodine. It does not result in a change to the hydrogen generation
analysis for containment. The increased mass inside containment will
have no significant impact on post-accident flood levels,
recirculation solution boron concentration, or peak clad
temperatures. No other operating parameters for systems, structures,
or components assumed to operate in the safety analysis are changed.
The offsite and control room doses meet the limits of 10 CFR [Part]
100 and GDC 19, respectively. Because the trisodium phosphate is
nonvolatile and the baskets are protected with solid covers and are
located slightly above the floor in the containment where access is
strictly controlled, a surveillance interval of once per refueling
outage provides assurance that the TSP will be available.
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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Project Director: Herbert N. Berkow.
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
[[Page 66719]]
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.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: June 3, 1996, as supplemented October
23, 1996.
Description of amendment request: The proposed amendment would
clarify a restriction on shutdown margin monitor operability while
changing modes so that it only limits reactivity changes caused by
boron dilution and rod withdrawal.
Date of publication of individual notice in Federal Register: June
20, 1996 (61 FR 31559).
Expiration date of individual notice: July 22, 1996.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385.
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.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: October 4, 1996 and
supplemented on November 6, 1996.
Brief description of amendments: The amendments add a Mode of
Applicability to Technical Specification 3.2.3.D, Inoperable Rod
Position Indicator Channels.
Date of issuance: November 25, 1996.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 176 and 163.
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 17, 1996 (61 FR
54240).
The November 6, 1996, submittal provided additional clarifying
information that did not affect the Commission's initial proposed
finding of no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 25, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: June 21, 1996.
Brief description of amendments: The amendments revise Technical
Specification (TS) Section 3/4.9.6, ``Manipulator Crane,'' to make the
wording consistent with the TS Bases description and consistent with
the design of the load handling equipment.
Date of issuance: November 25, 1996.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 156 and 148.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 23, 1996 (61 FR
55031) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 25, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: September 17, 1996 (TSC 96-01)
as supplemented October 23, 1996.
Brief description of amendments: The amendments lower the maximum
allowable reactor building pressure, lower the actuation setpoint for
actuation of the reactor building spray system, and modify the
associated TS Bases requirements.
Date of Issuance: November 25, 1996.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 219, 219, 216.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 23, 1996 (61 FR
55031). The October 23, 1996, letter provided clarifying information
that did not change the scope of the September 17, 1996, application
and the initial proposed no signficant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 25, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
[[Page 66720]]
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: July 31, 1996, as supplemented
by letters of September 5, October 22, and November 15, 20, and 21,
1996, which supersede the application submitted in the letter of May 9,
1996.
Brief description of amendment: The amendment (1) increased the
safety limit minimum critical power ratio (MCPR) for two loop operation
and single loop operation to 1.12 and 1.14, respectively, and (2) added
two General Electric topical reports to the list of documents
describing the analytical methods used to determine the core operating
limits. The changes are to Section 2.1.1, Reactor Core Safety Limits,
and Section 5.6.5, Core Operating Limits Report (COLR), respectively,
of the Technical Specifications. This amendment would go into effect in
Operating Cycle 9, at the end of the current Refueling Outage 8, and
the plant will have a mixed core of Siemens Power Corporation (SPS)
9 x 9-5 and General Electric (GE) GE11 reload fuel. The licensee also
changed the Bases of the Technical Specifications associated with the
above amendment.
Date of issuance: November 21, 1996.
Effective date: November 21, 1996.
Amendment No: 131.
Facility Operating License No. NPF-29: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 25, 1996.The
October 22, and November 15, 20, and 21, 1996, submittals provide
clarifying information that did not change the initial determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated November 21, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: April 15, 1996 (TSCR No. 244).
Brief description of amendment: The amendment revises Specification
5.3.1.B to allow the shield plug and the associated lifting hardware to
be moved over irradiated fuel assemblies that are in a dry shielded
canister within the transfer cask in the cask drop protection system.
Date of Issuance: November 7, 1996.
Effective date: November 7, 1996, to be implemented within 30 days
of issuance.
Amendment No.: 187.
Facility Operating License No. DPR-16: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20849). The Commission's related evaluation of this amendment and final
determination of no significant hazards consideration addressing
comments received on the proposed no significant hazards consideration
determination are contained in a Safety Evaluation dated November 7,
1996.
No significant hazards consideration comments received: Yes.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of application for amendment: February 22, 1996, and as
supplemented by letters dated July 24, October 4, November 19 and
November 25, 1996.
Brief description of amendment: The amendment changes Clinton Power
Station Technical Specification (TS) 3.3.8.1, ``Loss of Power
Instrumentation,'' and TS 3.8.1, ``AC Sources-Operating,'' by revising
the setpoint for the degraded voltage protection instrumentation and
modifying or deleting other Loss of Power Instrumentation TS
requirements. In addition, changes were also made to the minimum
required diesel generator voltage specified for certain diesel
generator surveillances.
Date of issuance: December 4, 1996.
Effective date: December 4, 1996.
Amendment No.: 110.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18168).The letters of July 24, October 4, November 19 and November 25,
1996, provided clarifying information and did not represent significant
changes from the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 4, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: October 11, 1996.
Brief description of amendments: These amendments revise Technical
Specification (TS) 3.9.6, ``Refueling Water Level,'' for San Onofre
Nuclear Generating Station (SONGS), Units 2 and 3. The proposed change
is required to restore certain provisions of the SONGS Units 2 and 3
operating practice that were not incorporated during the conversion to
the improved TS (Amendment Nos. 127 and 116, dated February 9, 1996).
Date of issuance: December 3, 1996.
Effective date: December 3, 1996, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2--134; Unit 3--123.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 1996 (61 FR
56251) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 3, 1996.
No significant hazards consideration comments received: No.
Temporary Local Public Document Room location: Science Library,
University of California, P.O. Box 19557, Irvine, California 92713.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 24, 1995, as supplemented by
letter dated July 26, 1996.
Brief description of amendment: The amendment revised Technical
Specification (TS) Surveillance Requirement 4.5.1.1.a.1 to base
accumulator operability on actual parameters (i.e., borated water
volume and nitrogen cover-pressure in the tanks) vs. the absence of
alarms.
Date of issuance: November 22, 1996.
Effective date: November 22, 1996, to be implemented within 30 days
of issuance.
Amendment No.: 103.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
[[Page 66721]]
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18632) The July 26, 1996, letter provided additional clarifying
information and did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated November 22, 1996.
No significant hazards consideration comments received: No.
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.
Dated at Rockville, Maryland, this 11th day of December 1996.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor
Regulation.
[FR Doc. 96-31944 Filed 12-17-96; 8:45 am]
BILLING CODE 7590-01-P
