[Federal Register Volume 62, Number 127 (Wednesday, July 2, 1997)]
[Notices]
[Pages 35846-35858]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-17140]
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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 June 9, 1997, through June 20, 1997. The
last biweekly notice was published on June 18, 1997 (62 FR 33117).
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 August 1, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714
[[Page 35847]]
which is available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, 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. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: May 6, 1997.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.7.5, ``Ultimate Heat Sink,''
and the associated bases to support steam generator replacement and to
incorporate recent Ultimate Heat Sink (UHS) design evaluations. The
replacement steam generators have a larger primary side volume which
results in a larger mass/energy release to the containment in the event
of a loss-of-coolant accident (LOCA), and a corresponding increase in
the heat load to the UHS.
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.
TS 3/4.7.5 establishes the operating requirements for the UHS.
Operation of the UHS within its design basis ensures the following:
(1) Sufficient cooling capacity is available for continued operation
of safety related equipment during normal and accident conditions
and (2) adequate inventory is available to provide a 30-day cooling
water supply to safety related equipment. Design analyses supporting
the proposed TS changes provide full qualification of the UHS.
A loss of off site power (LOOP) coincident with a loss of
coolant accident (LOCA), designated a LOOP/LOCA, on one unit, in
conjunction with the non-accident unit proceeding to an orderly
shutdown and cooldown from maximum power using normal operating
procedures, remains the limiting design basis event for the UHS
basin temperature.
The proposed changes to the UHS Limiting Condition for Operation
for basin temperature and the number of fans running do not, in
themselves, factor into any initiating event for Updated Final
Safety Analysis Report (UFSAR) Chapter 15 accidents and,
consequently, do not increase the probability of occurrence for
these previously evaluated accidents.
The UHS plays a vital role in mitigating the consequences of any
accident or transient. The proposed changes will ensure that the
[[Page 35848]]
minimum conditions necessary for the UHS to perform its design
functions will always be met. Engineering calculations demonstrate
that the SX [essential service water] pump discharge design
temperature limit of 100 deg.F, which was assumed as an initial
input for the accident analyses, is preserved. Consequently, the
proposed changes to the number of cooling tower fans required to be
running in high speed relative to the SX pump discharge temperature
do not increase the consequences of any accident previously
evaluated.
The two unit plant trip from full power with the loss of normal
auxiliary feedwater (AF) supply source has been shown to be more
limiting than the LOOP/LOCA scenario for UHS makeup and volume
considerations.
The proposed changes to the UHS LCO for minimum basin water
level do not, in themselves, factor into any initiating event for
the UFSAR Chapter 15 accidents and, consequently, do not increase
the probability of occurrence for these previously evaluated
accidents.
The proposed changes to increase the minimum basin water levels
ensure there is a sufficient volume of water in the UHS basin at all
times. With these proposed changes, the UHS will perform its design
function for the required 30 days, and the consequences of any
accident previously evaluated are not increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The supporting analyses for the revised TS 3/4.7.5 do not
involve a new or different kind of accident from any accident
previously evaluated. The proposed limits on SX basin minimum water
level, maximum basin temperature, and the number of fans operating
are within the design capabilities of the UHS, and ensure that the
UHS will always be in a condition to perform its design function in
the event of an accident or transient. New and revised analyses
which support the requested TS changes ensure the full qualification
of the UHS. The UHS will not be operated in a different manner such
that the possibility of a new or different kind of accident would be
created. Consequently, these changes do not create the possibility
of a new or different kind of accident from those previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed limits on SX basin minimum water level and maximum
temperature are based on the results of new and revised design
analyses which ensure that the margin of safety is not reduced.
Required operator actions with appropriate times are incorporated
into the analyses. The new limits on temperature and volume will
ensure that, under the most limiting accident or transient scenario,
cooling water from the basin will meet the accident analyses SX
design temperature limit of 100 degrees Fahrenheit and will ensure
that adequate inventory is available to provide a 30-day cooling
water supply to safety related equipment. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
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 amendment request: June 12, 1997.
Description of amendment request: The proposed license amendment
request would change the licensee's name from ``Duke Power Company'' to
``Duke Energy Corporation'' in the facility operating licenses for the
Catawba, McGuire, and Oconee nuclear stations as a result of a
corporate merger of Duke Power Company with PanEnergy Corporation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. These LARs (license amendment requests) involve an
administrative change only. The Oconee, McGuire, and Catawba FOLs
(Facility Operating Licenses) are being changed to reference the new
corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed changes.
Therefore, these LARs will have no impact on the possibility of any
type of accident: new, different, or previously evaluated.
(2) Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. These LARs involve an administrative change only. The
Oconee, McGuire, and Catawba FOLs are being changed to reference the
new corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed changes and no
failure modes not bounded by previously evaluated accidents will be
created. Therefore, these LARs will have no impact on the
possibility of any type of accident: new, different, or previously
evaluated.
(3) Will the change involve a significant reduction in a margin
of safety?
No. Margin of safety is associated with confidence in the
ability of the fission product barriers (i.e., fuel and fuel
cladding, Reactor Coolant System pressure boundary, and containment
structure) to limit the level of radiation dose to the public. These
LARs involve an administrative change only. The Oconee, McGuire, and
Catawba FOLs are being changed to reference the new corporate name
of the licensee.
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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North
Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: June 12, 1997
Description of amendment request: The proposed license amendment
request would change the licensee's name from ``Duke Power Company'' to
``Duke Energy Corporation'' in the facility operating licenses for the
Catawba, McGuire, and Oconee nuclear stations as a result of a
corporate merger of Duke Power Company with PanEnergy Corporation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. These LARs (license amendment requests) involve an
administrative change only. The Oconee, McGuire, and Catawba FOLs
(Facility Operating Licenses) are being changed to reference the new
corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed changes.
Therefore, these LARs will have no impact on the possibility of any
type of accident: new, different, or previously evaluated.
(2) Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
[[Page 35849]]
No. These LARs involve an administrative change only. The
Oconee, McGuire, and Catawba FOLs are being changed to reference the
new corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed changes and no
failure modes not bounded by previously evaluated accidents will be
created. Therefore, these LARs will have no impact on the
possibility of any type of accident: new, different, or previously
evaluated.
(3) Will the change involve a significant reduction in a margin
of safety?
No. Margin of safety is associated with confidence in the
ability of the fission product barriers (i.e., fuel and fuel
cladding, Reactor Coolant System pressure boundary, and containment
structure) to limit the level of radiation dose to the public. These
LARs involve an administrative change only.
The Oconee, McGuire, and Catawba FOLs are being changed to
reference the new corporate name of the licensee.
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: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, North Carolina 28223-0001.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: June 12, 1997.
Description of amendment request: The proposed license amendment
request would change the licensee's name from ``Duke Power Company'' to
``Duke Energy Corporation'' in the facility operating licenses for the
Catawba, McGuire, and Oconee nuclear stations as a result of a
corporate merger of Duke Power Company with PanEnergy Corporation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. These LARs (license amendment requests) involve an
administrative change only. The Oconee, McGuire, and Catawba FOLs
(Facility Operating Licenses) are being changed to reference the new
corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed changes.
Therefore, these LARs will have no impact on the possibility of any
type of accident: new, different, or previously evaluated.
(2) Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. These LARs involve an administrative change only. The
Oconee, McGuire, and Catawba FOLs are being changed to reference the
new corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed changes and no
failure modes not bounded by previously evaluated accidents will be
created. Therefore, these LARs will have no impact on the
possibility of any type of accident: new, different, or previously
evaluated.
(3) Will the change involve a significant reduction in a margin
of safety?
No. Margin of safety is associated with confidence in the
ability of the fission product barriers (i.e., fuel and fuel
cladding, Reactor Coolant System pressure boundary, and containment
structure) to limit the level of radiation dose to the public. These
LARs involve an administrative change only. The Oconee, McGuire, and
Catawba FOLs are being changed to reference the new corporate name
of the licensee.
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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
Attorney for licensee: J. Michael McGarry III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: May 30, 1997.
Description of amendment request: Technical Specification (TS)
Surveillances 4.5.2.f and 4.6.2.2.b require the periodic flow testing
of the recirculation spray system pumps. The proposed amendment would
change the surveillances by replacing the pump differential acceptance
criteria with a pump acceptance curve.
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:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Technical Specification Surveillances
4.5.2.f and 4.6.2.2.b will modify the surveillance acceptance
criteria to require that each Recirculation Spray System (RSS) pump
develop a differential pressure greater than or equal to the pump
performance curve contained on Figure 3.5-1 when tested according to
the requirements of Specification 4.0.5. Because it is undesirable
to test the pumps on recirculation flow to the RWST [reactor water
storage tank], pump testing will now be performed at lower flows
than previously performed. Consistent with Specification 4.0.5, one
point on Figure 3.5-1 will be used to meet the proposed surveillance
acceptance criteria. Periodically comparing the reference
differential pressure developed at this reduced flow detects trends
that might be indicative of pump degradation. The proposed changes
are consistent with RSS pump design criteria and performing
surveillance testing does not significantly increase the probability
of an accident previously evaluated.
The proposed changes to modify the surveillance acceptance
criteria to require that each RSS pump develop a differential
pressure greater than or equal to the pump performance curve
provides the necessary assurance that the pumps will function as
required in previous evaluations and does not significantly increase
the consequence of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to the surveillance acceptance criteria of
the RSS pumps does not change the operation of the Recirculation
Spray System or any of its components during normal or accident
evaluations.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will change the surveillance requirements
needed to demonstrate operability for each of the RSS pumps.
Technical Specification Surveillances 4.5.2.f and 4.6.2.2.b will now
require that each pump meet its acceptance criteria in accordance
with Figure 3.5-1
[[Page 35850]]
when tested according to the requirements of Specification 4.0.5.
Figure 3.5-1 will be inserted into the Technical Specifications.
The new acceptance criteria for the RSS Technical Specification
surveillance is above the accident analysis curve and is more
restrictive than the current inservice inspection curve in the
accident analysis region. The proposed TS curve has been degraded in
accordance with the recommendations of ASME XI (American Society of
Mechanical Engineers Boiler and Pressure Vessel Code, Section XI)
for the full range of flow and will be used to meet the TS
requirements.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 13, 1997.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) Surveillance Requirement 4.4.1.3.3
to be consistent with the requirements of TS 3.4.1.3. Specifically, the
change would bring TS Surveillance 4.4.1.3.3 into agreement with TS
3.4.1.3 that would require at least two reactor coolant system loops to
be operable and in operation when the reactor trip system breakers are
closed during Mode 4.
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:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve (an) SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed change to Technical Specification Surveillance
4.4.1.3.3 is being made to bring Technical Specification
Surveillance 4.4.1.3.3 into agreement with Technical Specification
3.4.1.3 that requires at least two reactor coolant system loops to
be operable and in operation when the reactor trip system breakers
are closed during Mode 4. This requirement was incorporated into
Technical Specification 3.4.1.3 in Amendment 7. This change to the
surveillance does not alter the design, operation, maintenance or
testing of the associated systems as previously analyzed.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This proposed change does not introduce any new failure modes or
malfunctions, since the changes only bring Surveillance 4.4.1.3.3 in
agreement with Technical Specification 3.4.1.3. Additionally, the
proposed change does not alter the operation of the reactor coolant
system during normal or accident conditions.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to Technical Specification Surveillance
4.4.1.3.3 will reword the surveillance to ensure compliance with
Technical Specification 3.4.1.3. Technical Specification 3.4.1.3 was
changed in Amendment No. 7 to address the closure of the Reactor
Trip System breakers in Mode 4. As written, Technical Specification
Surveillance 4.4.1.3.3 does not adequately ensure compliance with
Technical Specification 3.4.1.3. This proposed change is necessary
to bring Surveillance 4.4.1.3.3 in agreement with Technical
Specification 3.4.1.3 as it was amended.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of amendment requests: May 7, 1997, as supplemented May 30,
1997.
Description of amendment requests: The proposed amendments would
remove from the Technical Specifications certain limitations on crane
operations in the spent fuel pool enclosure relating to spent fuel pool
special ventilation system operability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Operation of the Prairie Island plant in accordance with the
proposed changes does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
proposed changes do not involve a physical modification to the
plant.
The spent fuel pool special ventilation system is provided to
mitigate the consequences of a design basis fuel handling accident
which involves dropping a spent fuel assembly directly onto a stored
spent fuel assembly. Spent fuel pool special ventilation system
performance and environmental consequences were based on the
conservative assumption that all fuel rods in one fuel assembly
fail. However, evaluation of the mechanical performance of spent
fuel stored in the spent fuel racks demonstrated that no fuel rods
fail.
The proposed changes will continue to require the spent fuel
pool special ventilation system to be operable to mitigate the
consequences of a fuel handling accident in accordance with its
original design intent. Spent fuel pool special ventilation system
operability is not required in conjunction with crane operations.
Heavy loads in the spent fuel pool enclosure are handled (1) by
single-failure-proof cranes with rigging and plant procedures which
implement Prairie Island commitments to NUREG-0612 [``Control of
Heavy Loads at Nuclear Power Plants''] or (2) over spent fuel pool
protective
[[Page 35851]]
covers as described in the Prairie Island USAR [updated safety
analysis report]. In accordance with the requirements of NUREG-0612,
use of a single-failure-proof crane with rigging and procedures
which implement the requirements of NUREG-0612 assures that the
potential for a load drop is extremely small and the effects of
heavy load drops are not considered. Spent fuel pool covers prevent
dropped loads from falling into the spent fuel pool. Thus, there are
no radiological releases resulting from handling heavy loads in the
spent fuel pool enclosure for which spent fuel pool special
ventilation system operability would be required. Therefore, these
changes do not involve a significant increase in the probability or
consequences of the fuel handling accident previously evaluated.
2. The proposed amendment(s) will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The proposed Technical Specification changes continue to require
the spent fuel pool special ventilation system to be operable during
handling of irradiated fuel as originally designed. Heavy loads in
the spent fuel pool enclosure are handled by means which assure that
the potential for a dropped load is extremely small (through use of
single-failure-proof cranes with rigging and plant procedures which
implement Prairie Island commitments to NUREG-0612) or prevent
dropped loads from falling into the spent fuel pool (through use of
spent fuel pool protective covers as described in the USAR). Thus,
the proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed changes, in themselves, do not introduce a new
mode of plant operation, surveillance requirement or involve a
physical modification to the plant.
The proposed changes do not alter the design, function, or
operation of any plant components and therefore, no new accident
scenarios are created. Therefore, the possibility of a new or
different kind of accident from any accident previously evaluated
would not be created by these amendments.
3. The proposed amendment(s) will not involve a significant
reduction in the margin of safety.
The proposed amendment(s) will continue to require the spent
fuel pool special ventilation system to operate following a fuel
handling accident as originally designed. Heavy load crane
operations in the spent fuel pool enclosure are handled (1) by
single-failure-proof cranes with rigging and plant procedures which
implement Prairie Island commitments to NUREG-0612; or (2) over
spent fuel pool protective covers as described in the Prairie Island
USAR. Provision of single-failure-proof equipment and compliance
with the other requirements of NUREG-0612 provides an equivalent
margin of safety to that which would be demonstrated by analysis of
the radiological effects of dropped loads. Use of protective covers
has been previously reviewed and approved by the NRC. Therefore,
th[ese] proposed amendment(s) (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 requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: John N. Hannon.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: May 9, 1997.
Description of amendment request: The proposed change revises the
Peach Bottom Atomic Power Station, Units 2 and 3 technical
specifications to extend the interval for replacing the primary
containment purge and exhaust valve inflatable seals.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS (technical specification) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Revising SR [surveillance requirement] 3.6.1.3.16 to replace the
inflatable seals for the Primary Containment purge and exhaust
valves from every 48 months to every 96 months will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The valves will continue to be leak
tight throughout the lifetime of the plant. This change will not
result in increased onsite or offsite radiological dose. This change
will result in reduced occupational dose exposure.
This submittal does not propose any change to the existing
requirements contained in the PBAPS [Peach Bottom Atomic Power
Station] Technical Specifications for leak testing of the Primary
Containment purge and exhaust valves per 10 CFR 50, Appendix J,
``Primary Reactor Containment Leakage Testing For Water-Cooled Power
Reactors.'' This continued testing will assure the leak tightness of
the purge and exhaust valves.
The T-ring materials (Ethylene Propylene) has been found to
withstand normal and accident thermal exposures for the design life
of the plant based on thermal aging analysis. The elastomer seat
material will provide acceptable seat tightness when exposed to a
total integrated radiation dose of 10E7 rads based on information
provided by EPRI [Electric Power Research Institute] in technical
report NP-2129, entitled ``Radiation Effects on Organic Material in
Nuclear Plants.'' The radiation dose of 10E7 rads bounds the design
basis accident dose to which these valves would be exposed. The
radiation dose these valves are exposed to during normal operation
is insignificant as compared to the accident dose. Based on this,
radiation effects from the additional exposure resulting from the
extended replacement frequency will not adversely impact the T-ring
seat material.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Revising SR 3.6.1.3.16 to replace the inflatable seals for the
Primary Containment purge and exhaust valves from every 48 months to
every 96 months does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This change does not involve any physical changes to a plant
structure, system, or component (SSC) which could act as an accident
initiator. The design, function, and reliability of the Primary
Containment purge and exhaust valves are also not impacted by this
change. This activity does not adversely influence any equipment,
which is required to be maintained operable for the prevention or
mitigation of accidents or transients. Furthermore, implementation
of the proposed changes will not adversely affect the manner in
which plant SSC are operated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
No margins of safety are reduced as a result of the proposed TS
changes. The proposed changes do not alter the intended operation of
plant structures, systems, or components utilized in the mitigation
of accidents or transients. The operating experience of these valves
and the testing performed in accordance with 10 CFR 50, Appendix J
provides a high level of confidence in the ability of these valves
to perform their intended safety function with respect to valve leak
tightness.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General
[[Page 35852]]
Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, PA
19101.
NRC Project Director: John F. Stolz.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: May 23, 1997.
Description of amendment request: The proposed change revises the
Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Technical
Specifications (TS) to exclude the measured Main Steam Isolation Valves
(MSIVs) leakage from the total Type B and C local leak rate test (LLRT)
results.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Excluding the MSIV leakage from the total Type B and C LLRT
results does not involve any change in the safety function or method
of operation of any plant component, system, or structure. No new
accident initiators or failure modes are created as a result of this
change. Therefore, this change will not result in an increase in the
probability of an accident previously evaluated.
The MSIV leakage release pathway is of significance only for the
evaluation of the design basis LOCA (loss-of-coolant accident) as
described in the PBAPS, Units 2 and 3 UFSAR (updated final safety
analysis report). The doses effectively reflected in the PBAPS,
Units 2 and 3 UFSAR reflect the impact of a 0.635% Primary
Containment volume per day Primary to Secondary Containment leakage,
plus a 0.145% Secondary Containment bypass leakage to the condenser.
Since accident consequences already reflect both leakage release
pathways, the consequences of the design basis LOCA are not
increased.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The MSIV's provide the means for mitigating the radiological
consequences of an accident. Revising Section 5.5.12 of the PBAPS,
Units 2 and 3 TS to exclude the measured MSIVs leakage from the
total Type B and C LLRT results has no effect on accident initiators
which lead to a new or different kind of accident. This change will
not involve any changes to plant systems, structures, or components
which could act as new accident initiators. The design, function,
and reliability of the MSIVs are also not impacted by this change.
Therefore, this change will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
No margins of safety are reduced as a result of this change to
the TS. No safety limits will be changed as a result of this TS
change. The MSIVs will continue to perform their intended safety
function. The combined dose rates from the two release paths (i.e.,
Primary to Secondary Containment leakage and Secondary Containment
bypass leakage) are unchanged as a result of this change, and are
within the limits of 10 CFR 100, and in conformance with NUREG-0737
post-accident access requirements.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: March 27, May 28, and June 4, 1997.
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) as follows:
Part 1--Boron Concentration Changes
The Cycle 2 core design for Watts Bar (WBN) will include a longer
fuel cycle and more highly enriched fuel (from 3.1 percent to 3.7
percent). To accommodate this design, the refueling water storage tank
(RWST) and accumulator boron concentrations will be increased to
provide enough boron in the sump to meet the large break loss-of-
coolant accident (LBLOCA) requirement for sump boron concentration.
This requirement is that during a LBLOCA, the core will remain
subcritical from boron provided by the emergency core cooling system
(ECCS), which takes suction from the RWST and containment sump.
The increase in RWST (TS 3.5.4) and accumulator (TS 3.5.1) boron
concentrations will be from a range of 2000-2100 ppm to 2500-2700 ppm
and from 1900-2100 to a range of 2400-2700 ppm, respectively.
Associated changes are proposed for TS Bases B 3.5.4.
Part 2--Safety Limits, Instrumentation, and Reactor Coolant System
Watts Bar has experienced hot leg temperature fluctuations,
including random spikes, which decrease the operating margin to both
the overtemperature delta temperature (OTDT) and overpower delta
temperature (OPDT) reactor trip setpoints. These fluctuations have
caused, in some cases, the plant to experience OT alarms during steady-
state operation since the temperature fluctuations reduced the
operating margin. To mitigate the temperature fluctuations and
associated alarms, the OTDT and OPDT setpoints have been enhanced to
increase the operating margin associated with these trip functions.
In addition, Watts Bar has decided to reduce the plant thermal
design flow from 97,500 gpm per loop to 93,100 gpm per loop (total of
390,000 gpm) to accommodate 10 percent steam generator tube plugging
and a 2 percent reduction in thermal design flow (RTDF).
Also, Watts Bar has decided to implement a tolerance of 0.6 deg.F
for the TS Surveillance for indicated differential temperature and 1
deg.F tolerance for the surveillance of TAVG (identified as
T prime and T double prime in the TSs). The use of this tolerance will
help to determine whether the indicated DT and TAVG should
be left as is, or rescaled during the surveillance. These tolerances
have been incorporated as biases into the uncertainty analysis for the
affected protection system functions. These functions include the OTDT,
OPDT and vessel DT equivalent to power (used in the steam generator
low-low water level trip functions). As a result of implementing these
biases into the protection system functions (and the changes to the
OTDT/OPDT setpoints and reduced TDF), the Allowable Value in the TSs
for the OTDT, OPDT and vessel DT equivalent to power functions have
been modified.
The licensee's safety evaluation has been prepared to allow for
plant operation during Cycle 2 with the revised OTDT and OPDT
setpoints, the thermal design flow of 93,100 gpm and the tolerances for
indicated differential temperature, T prime and T double prime. To
obtain sufficient departure from nucleate boiling (DNB) margin for the
OTDT/OPDT setpoint, reduced TDF and Cycle 2 design features, it was
necessary to implement the RTDP. The
[[Page 35853]]
RTDP program changes the uncertainty treatment for core power,
TAVG, pressurizer pressure, and RCS flow. These
uncertainties have been incorporated, where applicable, into the safety
analyses addressed in the Safety Evaluation.
The following TSs will be changed to incorporate the OTDT/OPDT
margin enhancement, thermal design flow of 93,100 gpm and tolerances
for indicated differential temperature, T prime and T double prime.
The Reactor Core Safety Limits (TS Figure 2.1.1-1 of the licensee's
application) have been modified to improve DNB margin. The Allowable
Values for the Vessel DT Equivalent to Power input to Steam Generator
Water Level Low-Low in the Reactor Trip System Instrumentation (Table
3.3.1-1, page 4) and Engineered Safety Feature Actuation System (ESFAS)
Instrumentation (Table 3.3.2-1, page 4), have been changed to reflect
the addition of a 0.6+F tolerance to the measurement of
indicated differential temperature.
The revised reactor core safety limits lines allow for changes in
the OTDT/OPDT reactor trip setpoints to improve operating margin. The
allowable values for these functions in the Reactor Trip System
Instrumentation (TS Table 3.3.1-1) have changed as a result of
including tolerances for indicated differential temperature, T prime
and T double prime in the uncertainty analysis. Several setpoint gains
and time constants have been modified to enhance plant operation.
Regarding the RCS Pressure, Temperature and Flow DNB Limits
(Section 3.4.1), the RCS average temperature limit has been revised to
account for the change in uncertainty from implementing RTDP. The total
RCS flow has been modified to account for the reduced thermal design
flow from 97,500 gpm to 93,100 gpm. The total flow value in the
Technical Specification includes an allowance for instrument
uncertainty.
Associated changes have been made to the following TS Bases
sections: Reactor Core Safety Limits (Section B 2.1.1); Nuclear
Enthalpy Rise Hot Channel Factor (Section B 3.2.2); Reactor Trip System
Functions OTDT, OPDT and Steam Generator Water Level Low-low (Vessel
Delta T Equivalent to Power) (Section B 3.3.1); Reactor Trip System
Functions--Reactor Coolant Flow--Low (Single Loop and Two Loops)
(Section B 3.3.1); ESFAS Instrumentation (Section B 3.3.2); RCS
Pressure, Temperature, and Flow DNB (Section B 3.4.1).
Part 3--Addition To Core Operating Limit Report Methodologies
The amendment would revise the Core Operating Limits Report (COLR)
methodologies listed in TS 5.9.5.b to add the reference to the
Westinghouse report WCAP-12610-P-A, ``Vantage + Fuel Assembly Reference
Core Report.'' The report reflects use of fuel assemblies in Cycle 2
using ZIRLO fuel rod cladding.
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:
Part 1--Boron Concentration Changes
The Nuclear Regulatory Commission has provided standards for
determining whether a significant hazards consideration exists (10 CFR
50.92 (c)). A proposed amendment to an operating license for a facility
involves no significant hazards consideration if 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;
The RWST and accumulator boron concentrations do not affect any
initiating event for accidents currently evaluated in the FSAR
[final safety analysis report]. The increased concentrations will
not adversely affect the performance of any system or component
which is placed in contact with the RWST or accumulator water. The
integrity and operability of the stainless steel surfaces in the
RWST, accumulator and affected NSSS [nuclear steam supply system]
components/systems will be maintained. The decrease in solution pH
is small and will not degrade the stainless steel. Also, the
integrity of the Class 1E instrumentation and control equipment will
be maintained since the lower sump pH, resulting from the increased
boron concentrations, is still within the applicable equipment
qualification [EQ] limits. These limits are set to preclude the
possibility of chloride induced stress corrosion cracking and assure
that there is no significant degradation of polymer materials. The
design, material and construction standards of all components which
are placed in contact with the RWST and accumulator water remain
unaffected.
For the evaluations, the consequences of an accident previously
evaluated in the FSAR will not be increased. There is no increase in
the LOCA accident consequences. The changes in the concentrations
increase the amount of boron in the sump during a LOCA. The
increased boron in the sump is sufficient to maintain the core in a
subcritical condition during a LOCA. Also, a revised hot leg
switchover time has been calculated and will be implemented in the
plant EOPs (emergency operating procedures). Thus, there will be no
boron precipitation in the core during a LOCA.
Furthermore, there is no increase in consequences of the non-
LOCA events. The concentration changes are a benefit to the SLB
(steam line break) at full power analysis due to the reduction in
power during the accident. The loss of normal feedwater event is not
sensitive to changes in the RWST and accumulator boron
concentrations. The concentration changes do not affect the
inadvertent operation of ECCS analysis since the minimum DNBR
(departure from nucleate boiling ratio) occurs at the event
initiation, and the concentration changes do not affect the analysis
trend.
Finally, the concentration changes are a benefit for the SLB M&E
(mass and energy) release and SGTR (steam generator tube rupture)
events since the increased boron increases the available shutdown
margin for these events. In addition, the increase in RWST and
accumulator boron concentrations and subsequent slight decrease in
containment sump and a spray pH does not impact the LOCA dose
evaluation since pH is not a function of radionuclide concentration.
Therefore, the present analysis remains bounding. Also, the slight
decrease in sump, core and spray fluid pH has been evaluated to not
impact the corrosion rate (and subsequent generation of Hydrogen) of
Aluminum and Zinc inside containment significantly that the present
analysis does not remain bounding. Further, the decreased sump, core
and spray fluid pH has been evaluated to not affect the amount of
hydrogen generated from the radiolytic decomposition of the sump and
core solution. In view of the preceding, it is concluded that the
proposed change will not increase the consequences of an accident
previously evaluated in the FSAR.
(2) or create the possibility of a new or different kind of
accident from any accident previously evaluated;
The changes to the RWST and accumulator concentrations do not
cause the initiation of any accident nor create any new credible
limiting single failure. The changes do not result in a condition
where the design, material, and construction standards of the RWST
and accumulators and other potentially affected NSSS components,
that were applicable prior to the changes, are altered. * * * *
The changes do not invalidate any of the accident analyses
results or conclusions. All of the safety analysis acceptance
criteria continue to be met. The changes in the concentrations
increase the amount of boron in the sump during a LOCA. The
increased boron in the sump is sufficient to maintain the core in a
subcritical condition during a LOCA. Also, a revised hot leg
switchover time has been calculated and will be implemented in the
plant EOPs. Thus, there will be no boron precipitation in the core
during a LOCA.
Furthermore, there is no possibility of a different kind of non-
LOCA event. The concentration changes are a benefit to the SLB at
full power analysis due to the reduction in power increase during
the accident. The loss of normal feedwater event is not sensitive to
changes in the RWST and
[[Page 35854]]
accumulator boron concentrations. The concentration changes do not
affect the inadvertent operation at ECCS analysis since the minimum
DNBR occurs at the event initiation, and the concentration changes
do not affect the analysis trend.
Finally, the concentration changes are a benefit for the SLB M&E
release and SGTR events since the increased boron increases the
available shutdown margin for these events.
(3) or involve a significant reduction in a margin of safety.
The changes do not invalidate any of the non-LOCA safety
analysis results or conclusions, and all of the non-LOCA safety
analysis acceptance criteria continue to be met. The margin of
safety associated with the licensing basis LBLOCA and SBLOCA (small-
break loss-of-coolant accident) analyses is not reduced as a result
of the proposed changes. Since adequate margin to the PCT (peak
cladding temperature) limit of 2200+F has been
maintained, no degradation in the margin of safety to the design
failure point (fuel melt) has been calculated. The licensing basis
containment and steam line break mass and energy releases remain
bounding, and the SGTR event acceptance criteria continue to be met.
Furthermore, the changes do not affect the safety related
performance of the RWST, accumulator or related NSSS components.
Part 2--Safety Limits, Instrumentation, and Reactor Coolant System.
The Nuclear Regulatory Commission has provided standards for
determining whether a significant hazards consideration exists (10
CFR 50.92 (c)). A proposed amendment to an operating license for a
facility involves no significant hazards consideration if 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;
The proposed changes do not result in a condition where the
design, material, and construction standards, which were applicable
prior to the changes, are altered. The revised OTDT and OPDT
setpoints do not require any hardware changes and are used for
accident mitigation. Thus, the setpoint changes do not increase the
probability of the accident.
All of the affected NSSS systems and components have been
evaluated with the TDF (thermal design flow) of 93,100 gpm. The
primary loop components (reactor vessel, reactor internals, CRDMs
(control rod drive mechanism), loop piping and supports, reactor
coolant pump, steam generator, and pressurizer) meet the applicable
structural limits with the revised TDF of 93,100 gpm and will
continue to perform their design functions. The RCCA (rod cluster
control assembly) drop time remains unaffected and the current
design core bypass flow remains valid. No additional steam generator
tubes need to be plugged to mitigate the potential for U-Bend
fatigue. Also, all of the NSSS systems will still perform their
intended design functions. The pressurizer spray flow remains above
the design value and the pressurizer relief system remains
unaffected since the TDF is lower than the current design flow and
the required pressure drop is lower. The design of the auxiliary
system components remains bounding for the revised TDF and the
corresponding changes to the NSSS thermal hydraulic parameters. In
addition, all of the NSSS/BOP (nuclear steam supply system/balance
of plant) interface systems will perform their intended design
functions. The steam generator safety valves will provide adequate
relief capacity to maintain the steam generator within applicable
design limits. The ADVs [atmospheric dump valves] will still relieve
20 percent of the maximum full load steam flow. The steam dump
system will still relieve 40 percent of the maximum full load steam
flow.
All of the applicable acceptance criteria for the accidents
described in the FSAR continue to be met. The LBLOCA analysis
currently uses a TDF of 93,100 gpm. Thus, no adjustments are
required for the LBLOCA input parameters to accommodate the TDF of
93,100 gpm. The SBLOCA has been performed with the TDF of 93,100
gpm, and the corresponding PCT is well below the 2200+F
limit. The post LOCA boron concentration and the hot leg switchover
time are unaffected. The revised thermal design procedure has been
implemented to obtain sufficient DNB margin to account for the TDF
of 93,100 gpm, the new OTDT/OPDT setpoints and the Cycle 2 design
features. All of the non-LOCA analyses have been re-analyzed or re-
evaluated and all of the applicable acceptance criteria continue to
be met.
The SLB radiological doses are unaffected and are still within
the existing licensing basis limits. The margin to overfill during
the SGTR event has been improved and the offsite doses during an
SGTR have been re-calculated and shown to be well within the
10CFR100 guidelines. The plant control systems will still provide
adequate response for the Condition 1 transients without causing a
reactor trip on OTDT and OPDT.
Finally, the changes in the tolerances for indicated
differential temperature, T prime and T double prime do not require
any hardware modifications and only require changes to the Technical
Specification Allowable Values for the OPDT and OTDT setpoints and
for the vessel DT equivalent to power functions. Thus, there is no
increase in the probability of an accident since the appropriate
Allowable Values have been modified to determine channel operability
for these functions.
(2) or create the possibility of a new or different kind of
accident from any accident previously evaluated;
The proposed changes do not cause the initiation of any accident
nor create any new limiting single failures. The OTDT and OPDT
protection functions are used for accident mitigation and do not
initiate any accidents. Also, the affected systems and components
will still perform their intended design functions.
* * *
The proposed changes do not create any new failure modes for
safety related equipment. The changes do not result in any original
design specification, such as seismic requirements, electrical
separation requirements or equipment qualification being altered.
The OTDT and OPDT setpoint changes do not require any hardware
modifications and only require adjustments to the setpoint values.
The setpoints are modeled in accident analyses which are used to
demonstrate equipment and structural qualification during a SLB.
With the setpoint changes and the TDF of 93,100 gpm, the current SLB
break M&E releases inside containment remain bounding and thus there
is no effect on the qualification of the equipment inside
containment during a SLB. The SLB M&E releases outside containment
have been re-calculated. The analysis of the impacts on equipment
qualification outside containment has been completed by generating
new temperature profiles. The application addresses and provides for
continued qualification of equipment through the normal EQ program.
Also, with the reduced TDF of 93,100 gpm, the current LOCA M&E
releases are still bounding, and thus there is no effect on the
qualification of equipment inside containment during a LOCA. The
OTDT and OPDT functions are not modeled in the LOCA analyses.
Furthermore, all of the applicable compartments and subcompartments
will maintain their integrity during the LOCA and the SLB since the
mass and energy releases for these compartments and subcompartments
remain unaffected.
In addition, the LOCA hydraulic forcing functions remain
bounding for the TDF of 93,100 gpm. Thus, the applicable NSSS
systems and components will still perform their structural functions
during a LOCA.
Finally, the changes in the tolerances for DTo, T
prime and T double prime do not require any hardware modifications
and only require changes to the Technical Specification Allowable
Values for the OPDT and OTDT setpoints and for the vessel DT
equivalent to power functions. Thus, there is no increase in the
probability of an accident different than any previously evaluated
since the appropriate Allowable Values have been modified to
determine channel operability for these functions.
(3) or involve a significant reduction in a margin of safety.
The margin of safety for the applicable safety analyses has not
been reduced. The OPDT and OTDT setpoints have been incorporated
into the affected safety analyses and all safety analysis criteria
continue to be met. All of the applicable DNB limits continue to be
met for the non-LOCA analyses. The LBLOCA input parameters do not
require adjustment for the TDF of 93,100 gpm. The SBLOCA has been
re-analyzed for the TDF of 93,100 gpm, and the SBLOCA PCT is well
below the 2200+F limit. The affected NSSS systems and
components will still meet the applicable design limits and perform
their intended safety functions with the TDF of 93,100 gpm. Also,
the SLB and LOCA M&E releases are still within the applicable
equipment qualification limits. The SGTR doses remain within the
applicable 10 CFR 100 limits, and the steam generator margin to
overfill is maintained.
Summary--Parts I and II. Based on the above, TVA has determined
that operation of
[[Page 35855]]
Watts Bar in accordance with the proposed amendment would not: (1)
involve a significant increase in the probability or consequences of
an accident previously evaluated, (2) create the possibility of a
new or different kind of accident from any accident previously
evaluated, or (3) involve a significant reduction in a margin of
safety. Therefore, operation of Watts Bar in accordance with the
proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92.
Part 3--Addition to Core Operating Limit Report Methodologies
(1) involve a significant increase in the probability or
consequences of an accident previously evaluated;
The use of ZIRLOTM is already permitted by TS section
4.2.1. Accordingly, the addition of the NRC approved Westinghouse
COLR methodology reference is administrative in nature. Therefore,
there is no increase in the probability or consequences of an
accident previously evaluated.
(2) or create the possibility of a new or different kind of
accident from any accident previously evaluated;
Since the use of ZIRLOTM is already permitted by TS
section 4.2.1, the addition of the NRC approved Westinghouse COLR
methodology reference is administrative in nature. Accordingly, no
new or different kind of accident has been created from those
previously evaluated.
(3) or involve a significant reduction in a margin of safety.
The use of ZIRLOTM is already permitted by TS section
4.2.1. The addition of the NRC approved Westinghouse COLR
methodology reference is administrative in nature. Therefore, there
is no significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 2, 1997.
Brief description of amendment request: The proposed amendment
would change the main steam isolation valve (MSIV) closure time
assumption used in the main steam line break accident analysis and
referenced in the Basis for Technical Specification 4.7.
Date of individual notice in Federal Register: May 15, 1997 (62 FR
26829).
Expiration date of individual notice: June 16, 1997.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: September 30, 1996, as supplemented
November 26, and December 12, 1996, February 13, March 5, April 2,
April 16, May 9, and June 3, 1997 (TSCR 192).
Description of amendment request: The proposed amendments would
change Technical Specification requirements related to the service
water system, component cooling water system, containment cooling and
iodine removal systems, auxiliary electrical systems, and the control
room emergency filtration system. The supplemental applications dated
April 2, April 16, May 9, and June 3, 1997, would eliminate separate
requirements for the component cooling water system for single-unit and
two-unit operation, revise the acceptance criteria for laboratory
testing of the control room emergency filtration system charcoal
adsorber banks from 90 percent to 99 percent, and supplement additional
information on the basis for acceptability of equipment qualification
analyses and dose assessments resulting from a loss-of-coolant
accident. The June 3, 1997, submittal requested the proposed amendments
be handled on an exigent basis based on the current schedule which
indicates that Unit 2 restart is scheduled for June 25, 1997, and Unit
1 restart is scheduled for July 1, 1997, and failure of the issuance of
the amendments by these dates would result in prevention of Point
Beach's resumption of operation.
Date of individual notice in the Federal Register: June 10, 1997
(62 FR 31636).
Expiration date of individual notice: July 10, 1997.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Notice of Issuance of Amendments to Facility Operating Licenses
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
[[Page 35856]]
local public document rooms for the particular facilities involved.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: June 20, 1996, as supplemented
by letters dated December 30, 1996, and March 5, 1997.
Brief description of amendments: The amendments would change the
Technical Specifications (TS) by incorporating NRC-approved thermal
limit licensing methodology in the list of approved methodologies used
in establishing the fuel cycle-specific thermal limits. In addition,
the proposed amendment will change the TS to reflect the use of Siemens
Power Corporation (SPC) ATRIUM-9B fuel for all operating Modes at
Dresden, Unit 3. The proposed amendment would also correct minor
editorial items in the TS.
Date of issuance: June 12, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 160 and 155.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the licenses and the Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17227). The Commission's related evaluation of the amendments is
cotained in a Safety Evaluation dated June 12, 1997.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 12, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.Consolidated
Edison Company of New York, Docket No. 50-247, Indian PointNuclear
Generating Unit No. 2, Westchester County, New York
Date of application for amendment: March 31, 1997.Brief description
of amendment: The amendment revises Technical Specifications (TSs) to
remove the reference of Valve 863 from TS Table 3.6-1. This revision
would allow for the installation of a proposed modification for
automatic closure of Valve 863 upon receipt of a Phase A containment
Isolation signal.
Date of issuance: June 19, 1997.
Effective date: As of the date of issuance to be implemented within 30
days.
Amendment No.: 193.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 15, 1997 (62 FR
26823)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 27, 1997, as supplemented by
letter dated May 6, 1997.
Brief description of amendment: The amendment changes the Technical
Specification 3/4.5.2, ``ECCS Subsystems--Modes 1, 2, and 3.'' The
proposed changes add a surveillance requirement to verify the Emergency
Core Cooling System (ECCS) piping is full of water at least once per 31
days, and clarifies wording of surveillance requirement 4.5.2.j. The
amendment also revises the TS Bases 3/4.5.2 and 3/4.5.3 to reflect
surveillance requirement.
Date of issuance: June 11, 1997.
Effective date: June 11, 1997, to be implemented within 60 days.
Amendment No.: 130.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17234). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 11, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida.
Date of application for amendment: December 20, 1996, and
supplemented February 13, and April 17, 1997.
Brief description of amendment: This amendment modifies the
Technical Specifications (TS) to delete a footnote associated with TS
2.1.1, ``Reactor Core Safety Limits'' which requires reactor thermal
power to be limited to 90% of 2700 Megawatts thermal for Cycle 14
operation beyond 7000 Effective Full Power Hours.
Date of Issuance: May 16, 1997.
Effective Date: May 16, 1997.
Amendment No.: 151.
Facility Operating License No. DPR-67: Amendment revised the TS.
Date of initial notice in Federal Register: January 15, 1997 (62 FR
2190).
The February 13, and April 17, 1997, letters provided clarifying
information that did not change the scope of the December 20, 1996,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 16, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: July 28, 1995, as revised
February 21, 1997.
Brief description of amendments: The amendments revise the
Technical Specifications for the Prairie Island Nuclear Generating
Plant to allow credit for soluble boron in spent fuel criticality
analyses. The request is based on the NRC approval of the Westinghouse
Owners Group generic methodology for crediting soluble boron given in
Topical Report WCAP-14416-NP-A, ``Westinghouse Spent Fuel Rack
Criticality Analysis Methodology,'' Revision 1, November 1996.
Date of issuance: June 12, 1997.
Effective date: June 12, 1997, with full implementation within 30
days.
Amendment Nos.: 129 and 121.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14464).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 12, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: January 13, 1997, as
supplemented March 24, 1997, May 13, 1997, and May 23, 1997.
[[Page 35857]]
Brief description of amendment: The amendment revises Technical
Specifications Requirements for containment leakage testing to add
several containment isolation valves and to implement the requirements
of 10 CFR Part 50, Appendix J, Option B for performance-based primary
reactor containment leakage testing.
Date of issuance: June 17, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 174.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 19, 1997 (62 FR
13173).
The March 24, May 13, and May 23, 1997, supplemental letters
provided clarifying information that did not change the initial
proposed no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 17, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: January 31, 1997.
Brief description of amendments: The amendments revise Technical
Specification 3/4.6.1.5, and its associated Bases section, to ensure
that a representative average containment air temperature is measured.
Date of issuance: June 13, 1996.
Effective date: Both units, as of the date of issuance, to be
implemented within 60 days.
Amendment Nos. 195 and 178.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 12, 1997 (62 FR
11497).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 13, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 22, 1996, as
supplemented March 28, 1997.
Brief description of amendments: Revise Technical Specifications
(TS) 3.6.5 and associated Bases to lower the minimum TS ice basket
weight. Also extend the chemical analysis surveillance interval for the
ice condenser ice bed from 12 months to 18 months.
Date of issuance: June 10, 1997.
Effective date: June 10, 1997.
Amendment Nos.: 224, 215.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19835).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: April 22, 1997, as supplemented
on May 15, and June 2, 1997. The April 22, 1997, submittal superseded a
previous submittal on this subject dated September 6, 1996 (61 FR
53769), as supplemented on October 30, October 31, November 7, November
15, and November 27, 1996, and January 23 and January 29, 1997.
Brief description of amendment: The amendment revises TS Section
4.2.b, ``Steam Generator Tubes,'' and its associated Basis, by allowing
a laser-welded repair of Westinghouse hybrid expansion joint (HEJ)
sleeved steam generator tubes.
Date of issuance: June 7, 1997.
Effective date: June 7, 1997.
Amendment No.: 135.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 7, 1997 (62 FR
24988).
The May 15, and June 2, 1997, submittals provided supplemental
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 7, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: April 24, 1997, as supplemented
on May 15 and 28, and June 5, 1997.
Brief description of amendment: The amendment revises TS Section
4.2.b, ``Steam Generator Tubes,'' to allow repair of steam generator
(SG) tubes with Combustion Engineering (CE) leak-tight sleeves in
accordance with CE generic topical report CEN-629-P, Revision 2,
``Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using
Leak-Tight Sleeves.'' The TS are also revised to allow re-sleeving of
tubes with existing sleeve joints in accordance with KNPP specific
topical report CEN-632-P, ``Repair of Kewaunee Steam Generator Tubes
Using a Re-Sleeving Technique.''
Date of issuance: June 7, 1997.
Effective date: June 7, 1997.
Amendment No.: 134.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 7, 1997 (62 FR
24989).
The May 15 and 28, and June 5, 1997, submittals provided
supplemental information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 7, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: April 28, 1997, as supplemented
on May 19, 1997.
Brief description of amendment: The amendment establishes a new
design basis flow rate for the auxiliary feedwater (AFW) pumps
consistent with the assumptions used in the reanalysis of the limiting
design basis event for the
[[Page 35858]]
AFW system. The Basis for TS 3.4.b, ``Auxiliary Feedwater System,'' has
been revised to reflect the change in AFW flow and to clarify the
requirements for the AFW cross-connect valves.
Date of issuance: June 7, 1997.
Effective date: June 7, 1997.
Amendment No.: 133.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 7, 1997 (62 FR
24977).
The May 19, 1997, submittal provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated June 7, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
Dated at Rockville, Maryland, this 25th day of June, 1997.
For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear
Reactor Regulation.
[FR Doc. 97-17140 Filed 7-1-97; 8:45 am]
BILLING CODE 7590-01-P