[Federal Register Volume 61, Number 14 (Monday, January 22, 1996)]
[Notices]
[Pages 1625-1647]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-676]
-----------------------------------------------------------------------
[[Page 1626]]
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 December 21, 1995, through January 4, 1996.
The last biweekly notice was published on January 3, 1996 (61 FR 174).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By February 21, 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 contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any
[[Page 1627]]
limitations in the order granting leave to intervene, and have the
opportunity to participate fully in the conduct of the hearing,
including the opportunity to present evidence and cross-examine
witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units
Nos. 1, 2, and 3, Maricopa County, Arizona.
Date of amendments request: December 19, 1995
Description of amendments request: The proposed amendments would
allow the implementation of the recently approved Option B to 10 CFR
Part 50, Appendix J. This new rule allows for a performance-based
option for determining the test frequency for containment leakage rate
testing. The proposed amendment would modify Technical Specifications
(TS) 1.7, 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, and 3/4.6.3 and the Bases of
TS 3/.6.1.2. It would also create a new TS 6.16.
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 Technical Specification (TS) changes will result in
generally increased intervals between containment leakage rate tests
determined through a performance based approach. The interval
between such tests are not related in any way to conditions which
cause accidents. Plant structures, systems, and components will not
be operated in a different manner as a result of the proposed TS
change, therefore, the proposed changes will not increase the
probability of an accident previously evaluated.
Containment leakage may result from accidents which are
evaluated in the Updated Final Safety Analysis Report. The proposed
TS changes may result in a small, but acceptable, increase in post-
accident containment leakage. This increase is calculated as a
statistical expectation using the probability that leakage through a
penetration will exceed the administrative limit and through the
increased time needed to detect such excess leakage. NUREG-1493,
which is the technical basis for 10 CFR Part 50, Appendix J, Option
B, contains a detailed evaluation of the expected leakage and its
consequences.
The increased risk due to the lengthening of the intervals
between Type A, B, and C leakage rate tests is also evaluated in
NUREG-1493. Using a statistical approach, NUREG-1493 determined that
the increase in expected dose to the public, resulting from
extending the testing interval, is extremely small. NUREG-1493
concluded that the small increase is justifiable due to the benefits
which accrue from interval extension. The primary benefit is the
reduction in occupational exposure. The reduction, on a per person
basis, is orders of magnitude greater than the marginal, potential
increase in dose to the public. The reduction in occupational
exposure is a real reduction, while the small increase in dose to
the public is statistically derived using conservative assumptions.
Therefore, the proposed change does not significantly increase the
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated. The
proposed change only incorporates the performance based approach
authorized in the new Option B to Appendix J of 10 CFR Part 50. The
interval extensions allowed, through this approach, do not have the
potential for creating the possibility of new or different kinds of
accidents from those previously evaluated. Plant structures,
systems, and components will not be operated in a different manner
as a result of the TS change and, therefore, will not introduce any
new or different failure modes or initiators.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed Technical Specification does not alter the
allowable containment leakage rate. The proposed change replaces the
current, prescriptive testing requirements with a new performance
based approach for establishing the testing intervals therefore, the
proposed change does not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland.
Date of amendment request: December 21, 1995.
Description of amendment request: The proposed amendment would
revise the Calvert Cliffs Nuclear Power Plant,
[[Page 1628]]
Unit No. 1, Technical Specifications (TSs). The requested change would
allow the use of cladding materials other than Zircaloy or ZIRLO. A
Temporary Exemption was issued on November 28, 1995 (60 FR 62483)
approving the loading of four (4) lead fuel assemblies (LFAs) into the
Unit No. 1 reactor vessel during cycles 13, 14, and 15. The technical
basis for the Exemption, which is the same basis for the requested TS
amendment, was provided in the Baltimore Gas and Electric Company (BGE)
submittal dated July 13, 1995. The submittal addressed the safety
significance of operating with 4 LFAs in Calvert Cliffs Nuclear Power
Plant, Unit No. 1, reactor vessel during cycles 13, 14, and 15.
Specifically, BGE proposes to add a statement to TS 5.2.1, ``Fuel
Assemblies,'' indicating, for Cycles 13, 14, and 15 only, advanced
cladding material may be used in 4 lead test assemblies as described in
a approved Temporary Exemption dated November 28, 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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change is to add an approved temporary exemption to
the Unit 1 Technical Specifications allowing the installation of
four lead fuel assemblies. These four assemblies use an advanced
cladding material which is not specifically permitted by existing
regulations or Calvert Cliffs' Technical Specifications. A temporary
exemption to allow the installation of these assemblies was approved
on November 28, 1995. The addition of this approved temporary
exemption to Technical Specification 5.2.1 is simply intended to
allow their installation under the provisions of the temporary
exemption. The license amendment is effective only as long as the
exemption is effective. The addition of the approved temporary
exemption to Unit 1 Technical Specification 5.2.1 does not change
the probability or consequences of an accident previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed Technical Specification change adds an approved
temporary exemption to Technical Specification 5.2.1 for Unit 1.
This change does not add any new equipment, modify any interfaces
with existing equipment, change the equipment's function, or change
the method of operating the equipment. The proposed change does not
affect normal plant operations or configuration. Since the proposed
change does not change the design, configuration, or operation, it
could not become an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The proposed change is to add an approved temporary exemption to
the Unit 1 Technical Specifications allowing the installation of
four lead fuel assemblies. These four assemblies use an advanced
cladding material which is not specifically permitted by existing
regulations or Calvert Cliffs' Technical Specifications. A temporary
exemption to allow the installation of these assemblies was approved
on November 28, 1995. The addition of this approved temporary
exemption to Technical Specification 5.2.1 is simply intended to
allow their installation under the provisions of the temporary
exemption. The license amendment is effective only as long as the
exemption is effective. This amendment does not change the margin of
safety by adding a reference to an approved, temporary exemption to
the Technical Specifications.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh.
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: December 7, 1995.
Description of amendment request: The proposed amendments will
remove the Technical Specification (TS) requirements for the main
feedwater pump discharge pressure switch input to the Anticipatory
Reactor Trip System (ARTS) and the Emergency Feedwater System (EFDW).
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:
No. The accidents addressed within the Oconee Final Safety
Analysis Report (FSAR) have been reviewed with respect to this
proposed Technical Specification amendment request. The probability
or consequences of any accident previously evaluated is not
significantly increased by the proposed amendment. Emergency
Feedwater is required for the mitigation of some accidents and the
availability of this system will be unaffected by this proposed
revision. Both manual and automatic actuation of the EFDW system on
a loss of main feedwater will remain.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
No. This amendment eliminates a portion of the automatic
actuation circuitry for EFDW and ARTS. This circuitry removal does
not create the possibility of a new or different kind of accident as
the design of the circuitry is to sense a loss of main feedwater and
supply a signal for the initiation of ARTS and EFDW. A loss of main
feedwater signal will continue to be supplied to ARTS and EFDW;
however, this loss will be sensed by low hydraulic oil pressure on
the Main Feedwater Pumps (ARTS and EFDW) and low steam generator
level (EFDW only) rather than by a low Main Feedwater Pump discharge
pressure. Since a loss of Main Feedwater will continue to be
recognized, the system will continue to function as before. Hence,
no new or different accidents will be created.
(3) Involve a significant reduction in a margin of safety.
No. The margin of safety will not be significantly reduced as an
actuation signal to ARTS and EFDW will continue to be generated by a
loss of Main Feedwater. Consequently, ARTS and EFDW will continue to
perform the safety function required for accident mitigation.
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.
[[Page 1629]]
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.
Date of amendment request: November 22, 1995.
Description of amendment request: The proposed amendments will
upgrade existing TS [Technical Specification] 3/4.4.6.1 for the Reactor
Coolant System Leakage Detection Instrumentation by adapting the
Standard Technical Specifications for Combustion Engineering Plants
(NUREG-1432), Specification 3.4.15, to both St. Lucie units. The
proposal is consistent with the NRC Final Policy Statement on Technical
Specifications Improvements (58 FR 39132).
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 Reactor Coolant System (RCS) Leakage Detection
Instrumentation Systems are not accident initiators, and their
operational status is not a consideration in determining the
probability of occurrence of accidents previously evaluated. The
proposed revision to the related Limiting Condition for Operation
(LCO) 3/4.4.6.1 does not involve a change to the configuration or
method of operation of any equipment that is used to mitigate the
consequences of an accident, nor do the changes alter any
assumptions made involving initial plant conditions in the safety
analyses. Therefore, operation of the facility in accordance with
the proposed amendment would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed revision to LCO 3/4.4.6.1 is administrative in
nature and will not result in a change to the physical plant or the
modes of plant operation defined in the Facility License. The
revision does not involve the addition or modification of equipment
nor does it alter the design of plant systems. Therefore, operation
of the facility in accordance with the proposed amendment would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The RCS Leakage Detection Systems are designed to provide
diverse methods to assist in the detection and location of
unidentified leakage that may be associated with potential pressure
boundary degradation. These systems provide no equipment control or
accident mitigation functions, and are not associated with the
safety margin established for protection from analyzed Loss of
Coolant Accidents. The proposed revision to LCO 3/4.4.6.1 does not
alter the basis for any technical specification that is related to
the establishment of, or the maintenance of, a nuclear safety
margin; and simply adapts the corresponding and previously reviewed
specification from the Standard Technical Specifications for
Combustion Engineering Plants, NUREG-1432, to the St. Lucie units.
Therefore, operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Based on the above discussions and the supporting Evaluation of
Technical Specification changes, FPL has determined that the
proposed license amendment involves no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: David B. Matthews, Director.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey.
Date of amendment request: December 5, 1995.
Description of amendment request: The proposed amendment revises
the submittal date in the Annual Exposure Data Report which brings
Oyster Creek into conformance with 10 CFR 20.2206 and relaxes an overly
restrictive administrative requirement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
. . . The changes do not:
1. Involve a significant increase in the probability or the
consequence of an accident previously evaluated.
This change is administrative in nature and has no effect on the
operation of the plant. This change will not increase the
probability or consequence of an accident previously evaluated.
2. Create the possibility a new or different kind of accident
from any previously evaluated.
Operation of the facility in accordance with this proposed
change will not create the possibility for an accident or
malfunction of a different type from any accident previously
evaluated. The proposed amendment does not modify any system
(component) operation or maintenance activity. The facility will
continue to be operated within the limits of existing accident
analysis and margins of safety.
3. Involve a significant reduction in a margin of safety.
This change brings the submittal date for the Annual Exposure
Data Report into conformance with 10 CFR 20.2206 and relaxes an
overly restrictive administrative requirement. Since the proposed
change does not alter any system hardware or design basis, 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: Phillip F. McKee.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa.
Date of amendment request: November 15, 1995.
Description of amendment request: The proposed amendment would
revise the requirements for the End of Cycle Recirculation Pump Trip
logic to match more closely the assumptions applicable to the turbine
trip events for which it was installed. The surveillance requirements
are also proposed to be revised, based on those same assumptions.
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 Specification (TS) amendment will not
significantly increase the probability or consequences of any
previously evaluated accidents. The [End of Cycle] (EOC)
[recirculation pump trip] RPT system was installed to preclude
[[Page 1630]]
violation of reactor fuel limits, and the system will be preserved for
that purpose. In the event that system is not available, an
operating penalty will be imposed on the [Minimum Critical Power
Ratio] MCPR limit to assure sufficient margin to the limit to
preclude fuel damage during the postulated turbine trip events.
The change to the ``Minimum Operable Channels per Trip System''
will assure that inputs monitoring both the turbine control valve
fast closure and the turbine stop valve closure will be available to
initiate (EOC)RPT.
The change to the ``Applicable Operating Mode'' is an editorial
change which reflects the existing hardware bypass.
The change to Action 81 in TS Table 3.2-G will assure that when
the (EOC)RPT system does not meet the minimum TS availability
requirements, the [safety limit minimum critical power ratio] SLMCPR
will not be challenged. By imposing an [operating limit minimum core
power ratio] OLMCPR penalty for continued operation, the fuel
thermal limits will not be challenged, since the (EOC)RPT system was
installed to accomplish the same goal. No increase in the
consequences of the turbine trip events will result from this
change. The OLMCPR penalty is dependent on cycle-specific parameters
and will therefore be included in the cycle-specific [Core Operating
Limits Report] COLR.
The change to the surveillance interval results in (EOC)RPT
logic channel functional tests being performed once per quarter
instead of once per month. The change also revises the allowed out-
of-service time (AOT) for testing from two hours to six hours. These
changes are consistent with the Improved Standard Technical
Specifications, NUREG-1433, Revision 1. The (EOC)RPT is initiated by
instruments common to the Reactor Protection System (RPS) (i.e.,
turbine stop valve closure and turbine control valve fast closure).
The surveillance interval and AOT changes for these instruments were
evaluated in ``Technical Specification Improvement Analysis for BWR
Reactor Protection System,'' NEDC-30851P-A, March 1988, for the RPS
function. Although the (EOC)RPT functions were not explicitly
identified in that document, these changes can be considered bounded
by that analysis. The basis for this conclusion is similar to the
basis established for the control rod block instrumentation common
to the RPS, as documented in ``Technical Specification Improvement
analysis for BWR Control Rod Block Instrumentation,'' NEDC-30851P-A,
Supplement 1, October 1988. Failure of the (EOC)RPT function could
potentially lead to exceeding the SLMCPR, similar to the
consequences of an unmitigated rod withdrawal error. The slight
increase in risk of a SLMCPR violation due to extending (EOC)RPT
surveillance interval and AOT is offset by the same benefits
associated with the similar approved surveillance interval and AOT
for the RPS. Both the above referenced reports have been approved
for application at the DAEC via TS Amendment 193, dated April 14,
1993.
The changes to the ``Operating Modes for which Surveillance
Required'' are clarifications and will result in a more efficient
utilization of resources. By stating that the surveillance applies
only when the (EOC)RPT system is OPERABLE, the surveillances will
not be performed needlessly. During the early part of an OPERATING
cycle, the (EOC)RPT is not required to mitigate a turbine trip, and
therefore, may be bypassed. At the time when the (EOC)RPT is assumed
to be OPERABLE pursuant to the analysis, it will be made OPERABLE
unless accepting the penalty on the OLMCPR is preferable. The result
of the proposed change will still be that the (EOC)RPT is
demonstrated OPERABLE at any time when it is required.
The change to the acceptance criteria for response time testing
reflects a recent review of the analytical assumptions and the
testing methodology. The (EOC)RPT is assumed to interrupt power to
the recirculation pump motor within 175 milliseconds after
initiation of either turbine stop valve closure or turbine control
valve fast closure. The response time test only measures a portion
of the complete trip (the rest was measured as part of start-up
testing). The portion measured is dependent on which trip input is
being tested. The turbine control valve closure is sensed by a
pressure switch monitoring the hydraulic fluid controlling the valve
and therefore has no delay between valve motion and initiation of
the (EOC)RPT logic. The turbine stop valve closure is sensed by
position switch. Since this switch is set to initiate (EOC)RPT at
10% valve closed, there is a brief delay between the beginning of
valve motion and initiation of the (EOC)RPT logic. The respective
proposed response time tests account for these differences, as
described in the footnotes on TS page 3.2-36, and demonstrate that
the measured portions of the action are within allowed time periods.
None of the proposed changes will significantly increase the
probability of any accident previously evaluated because the
(EOC)RPT is not an initiator of any of those events. None of the
proposed changes will significantly increase the consequences of an
accident because the (EOC)RPT system serves to prevent a turbine
trip event from exceeding the fuel SLMCPR, and it will continue to
perform in that capacity at any time when it is required to assure
margin to the SLMCPR.
2. The proposed changes will not add a new or different kind of
accident because the plant will not be operated in a different way.
By allowing the implementation of a penalty on OLMCPR in lieu of
reducing reactor power, the risk of a plant transient is reduced.
Similarly, the surveillance interval and AOT extensions will also
result in fewer plant power reductions for testing.
The (EOC)RPT initiates a trip of the recirculation pumps and any
TS change affecting that system cannot result in an effect on any
system other than those pumps. Consequently, no new accidents are
postulated as a result of this proposed change.
3. The proposed change will not result in a significant
reduction in any margin of safety. The (EOC)RPT performs to assure
adequate margin to the SLMCPR. The proposed change will preserve
that function and require that additional margin to the SLMCPR be
imposed for those times when the (EOC)RPT is not OPERABLE. The other
changes are proposed because they assure correct (EOC)RPT function
(inputs and response times).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: Gail H. Marcus.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois.
Date of amendment request: December 14, 1995.
Description of amendment request: The proposed amendment would
modify Technical Specification 3.4.2, ``Flow Control Valves (FCVs),''
by deleting the requirement to verify that the average rate of movement
of each reactor recirculation system FCV is limited to less than or
equal to 11% per second in the opening and closing directions
(Surveillance Requirement 3.4.2.2).
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 Clinton Power Station (CPS) Updated Safety Analysis
Report (USAR) evaluates three specific events related to operation
of the reactor recirculation flow control valves (FCVs). The impact
of the proposed change on each of these events is discussed below.
The loss of coolant accident (LOCA) analysis described in USAR
Section 6.3.3.7.2 assumes that the FCVs fail ``as is'' in the event
of a LOCA. This feature is assured by electronic interlocks in the
FCV control circuitry and periodically verified as required by
Technical Specification (TS) Surveillance Requirement (SR) 3.4.2.1.
The design of these interlocks and the testing requirements are not
affected by this proposed change.
The Recirculation Flow Controller Failure--Decreasing Flow
transient analyses are described in USAR Section 15.3.2, and the
Recirculation Flow Controller Failure--Increasing Flow transient
analyses are described in USAR Section 15.4.5. Since the
[[Page 1631]]
control circuitry for the FCVs has been modified such that the
capability to operate in a master controller mode has been
eliminated, each FCV is now individually controlled, and the
possibility that a single failure could affect operation of more
than one FCV has also been eliminated. As a result, fact closure and
fast opening of both FCVs are no longer postulated for CPS. Thus,
the surveillance (SR 3.4.2.2) associated with verifying that FCV
movement is within the assumptions of the analyses for fast closure
and fast opening of both FCVs can be deleted.
With respect to fast closure and fast opening of individual
FCVs, the modification performed during the fifth refueling outage
only affected the electronic master control of the FCVs and did not
affect the hydraulic limitations of the FCVs. Conservative analyses,
component testing, and the Initial Startup Test program provide
confidence that individual FCV stroke rates assumed in the transient
analyses will not be exceeded over the life of the plant. These
analyses and conditions are sufficient to assure individual FCV
stroke rates are adequately limited without the periodic performance
of a specific test.
In addition to the above, the modification did not add any new
failure modes to the design of the individual FCV controllers. In
fact, failure modes associated with misoperation of the common
master controller have been eliminated from the control circuit
design. The modification did not alter any of the features
associated with initiators of any LOCA or features which assure that
the FCVs fail ``as is'' in the event of a LOCA.
Based on the above, Illinois Power (IP) has concluded that this
request does not increase the probability or the consequences of any
accident (or transient) previously evaluated.
(2) USAR Sections 15.3.2 and 15.4.5 describe the plant response
to malfunctions of FCV control failures, and USAR Section 6.3.3.7.2
describes the assumptions made with respect to FCV failures and
their impact on the LOCA analysis. The proposed change (and the
associated modification prompting the proposed change) does not
affect any other structures, systems, or components beyond the FCVs.
All associated failure modes thus remain within the scope of the
failure modes previously considered. As a result, IP has concluded
that the proposed change cannot create the possibility of an
accident not previously evaluated.
(3) This request does not involve any change to the requirements
or design associated with initiation or mitigation of a LOCA. The
consequences of transients associated with fast closure and fast
opening of reactor recirculation system FCVs are bounded by the
consequences of other transient events and thus are not utilized in
establishing plant operating limits. Although the control circuitry
for the FCVs was modified during the fifth refueling outage, that
modification did not affect the hydraulic failure modes of the FCVs.
Further, the modification did not add any new failure modes to the
design of the individual FCV controllers. In fact, failure modes
associated with misoperation of the common master controller have
been eliminated from the control circuit design. As a result,
assumed FCV operation during analyzed accidents and transients has
not been altered. Conservative analysis, component testing, and the
Initial Startup Testing program have confirmed that the FCV velocity
assumed in the transient analyses will not be exceeded over the life
of the plant. Thus, verification of rate of FCV movement in the
opening and closing directions need not be performed by periodic
testing and SR 3.4.2.2 can be deleted without resulting in a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606.
NRC Project Director: Gail H. Marcus.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois.
Date of amendment request: December 14, 1995.
Description of amendment request: The proposed amendment would
consist of several changes to the instrumentation sections of the
Clinton Power Station Technical Specifications. The proposed changes
are required due to engineering reanalyses or plant modifications. The
affected instrumentation includes: (1) steam line flow high channels
for the Reactor Core Isolation Cooling (RCIC) System, (2) ambient
temperature channels in the Residual Heat Removal (RHR) System heat
exchanger rooms, (3) reactor vessel pressure channels that provide a
permissive for operation of the shutdown cooling mode of the RHR
system, and (4) RCIC storage tank water level instrument channels.
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) None of the proposed changes involve a significant increase
in the probability or consequences of any accident previously
evaluated.
The changes to Table 3.3.6.1-1 Functions 3.a and 3.i are
administrative in nature and bring the technical specifications (TS)
into conformance with the Clinton Power Station (CPS) as-built
design. The reactor core isolation cooling (RCIC) system steam line
flow trip Function names have been changed to reflect the
elimination of the residual heat removal (RHR) steam condensing
mode. However, these trips have not been physically altered and thus
will continue to operate as before. As a result of the elimination
of the RHR steam condensing mode, the possibility of a leak in the
RCIC steam supply resulting in an increase in the RHR heat exchanger
room ambient temperature has also been eliminated. Accordingly, the
RHR ambient temperature isolation trip is changed to only isolate
the RHR system when the RHR heat exchanger room ambient temperature
setpoint is exceeded. The Shutdown Cooling System Reactor Vessel
Pressure--High function is provided to isolate the shutdown cooling
portion of the RHR system since this piping is designed for
pressures lower than rated reactor vessel pressure. This interlock
(RHR cut in permissive) is provided only for equipment protection to
prevent an intersystem LOCA scenario and credit for the interlock is
not assumed in the accident or transient analysis in the Updated
Safety Analysis Report (USAR).
The proposed change to the setpoint (Allowable Value) is
conservative with respect to considerations for shutting the RHR
shutdown cooling motor-operated valves and providing
overpressurization protection for the low pressure RHR shutdown
cooling system piping. With respect to the RCIC storage tank water
level setpoints, no accident or transient analysis takes credit for
the volume of water in the RCIC storage tank. In addition, the
setpoint (Allowable Value) has been changed to ensure RCIC system
operation is not adversely affected by a low level in the storage
tank.
The proposed changes do not affect any of the parameters or
conditions that contribute to initiation of any accidents previously
evaluated. In addition, the proposed changes do not affect the
ability of the associated instrumentation to operate as assumed in
the safety analyses. As a result, the proposed changes will not
result in a significant increase in the consequences of any accident
previously evaluated.
(2) None of the proposed changes create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed changes for RHR/RCIC Steam Line Flow--High
[are] administrative in nature and will simply make this item
description accurate. The RCIC steam supply line no longer supplies
any steam to the RHR heat exchanger room. As a result, the
associated isolation of the RCIC system is no longer required. The
Shutdown Cooling System Reactor Vessel Pressure - High function will
still perform as designed. The RCIC Storage Tank Level - Low trip
will continue to perform in accordance with design. None of the
above listed changes will introduce any new failure modes or changes
in plant operation.
As a result, the proposed changes cannot create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) None of the proposed changes involve a significant reduction
in a margin to safety.
[[Page 1632]]
The proposed changes for RHR/RCIC Steam Line Flow--High do not involve
a significant reduction in a margin of safety because the change is
administrative in nature and will simply make the descriptions
accurate and consistent with completed modifications. The
elimination of RCIC system isolation in response to a high RHR room
ambient temperature is no longer required due to the elimination of
the RHR steam condensing mode. Removing the RHR room ambient
temperature isolation of the RCIC will reduce the number of
unnecessary isolations of RCIC. The Shutdown Cooling System Reactor
Vessel Pressure - High function will still perform as designed. The
proposed change to the setpoint (Allowable Value) is conservative
with respect to considerations for shutting the RHR shutdown cooling
motor-operated valves and providing overpressurization protection
for the low pressure RHR shutdown cooling system piping. The
Allowable Value for the RCIC Storage Tank Level - Low Function has
been changed to be more conservative to ensure the RCIC and HPCS
systems will perform their system safety function. No credit is
taken for the volume in the RCIC storage tank for the HPCS or RCIC
systems in performing their safety-related functions.
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: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606
NRC Project Director: Gail H. Marcus.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan.
Date of amendment requests: December 19, 1995 [AEP:NRC:1215B]
Description of amendment requests: The proposed amendments would
modify the technical specifications to replace the existing scheduling
requirements for overall integrated and local containment leakage rate
testing with a requirement to perform the testing in accordance with 10
CFR Part 50, Appendix J, Option B. Option B allows test scheduling to
be adjusted based on past performance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
This amendment request does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the proposed changes to the T/Ss do not affect the
assumptions, parameters, or results of any UFSAR [updated final
safety analysis report] accident analysis. The proposed changes do
not change the acceptance criteria for containment leakage limits
and do not modify the response of the containment during a design
basis accident. The proposed amendment does not add or modify any
existing equipment. The proposed Types A, B, and C testing schedules
will be consistent with Appendix J Option B to 10 CFR 50 which was
developed based on analytical efforts documented in NUREG-1493
[Performance-Based Containment Leak-Test Program]. The analysis
confirms previous observations of insensitivity of population risks
from severe reactor accidents to containment leakage rates. Based on
these considerations, it is concluded that the changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Criterion 2
The proposed changes do not involve physical changes to the
plant or changes in plant operating configuration. The proposed
changes only remove the restrictive schedular requirements for
conducting Types A, B, and C testing from the T/Ss and substitute
the schedule specified in Appendix J Option B to 10 CFR 50 and
Regulatory Guide 1.163 [Performance-Based Containment Leak-Test
Program]. Thus, it is concluded that the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Criterion 3
Based on NUREG-1493, Regulatory Guide 1.163, and the rule
posting in the Federal Register (60 FR 49495), the margin for safety
presently provided is not significantly reduced by the proposed
change to a performance-based test interval for Types A, B, and C
tests. Although the changes allow more flexibility in scheduling
tests, the proposed amendment continues to ensure reactor
containment system reliability by periodic testing in full
compliance with 10 CFR 50, Appendix J Option B. Based on these
considerations, it is concluded that the changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: John N. Hannon.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota.
Date of amendment request: August 15, 1995, as supplemented
November 14, 1995.
Description of amendment request: The proposed amendment would
modify the Monticello Technical Specifications (TS) to: (1) revise the
main steam line isolation valve leak rate test acceptance criterion to
be based upon the combined maximum flow path leakage for all four main
steam lines of 46 standard cubic feet per hour (scfh) in lieu of the
current limit of 11.5 scfh per valve; (2) revise the operability test
interval for the drywell spray header and nozzles from 5 years to 10
years; and (3) revise TS 3/4.7.a.2, Primary Containment Integrity, to
remove information specific to the primary containment leakage rate
testing program and replace it with a commitment to abide by the
requirements of 10 CFR Part 50, Appendix J, Option B, Section III.A,
for Type A testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
a. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment is limited to changes to the surveillance
testing requirements applicable to the main steam line isolation
valves [MSIVs] allowable leakage criteria, drywell nozzles test
interval, and method of applying Appendix J test requirements. With
respect to monitoring main steam [line] isolation valve performance,
the proposed criteria are equivalent to the current criteria
ensuring that leakage past the valves would be within acceptable
limits under accident conditions. These surveillance tests are
performed while the plant is in a cold shutdown condition at a time
when the equipment is not required to be operable. Performance of
the tests themselves are not input or consideration in any accident
previously evaluated, thus the proposed change will not increase the
probability of any such accident occurring.
The proposed amendment will not adversely affect the function,
operation, or reliability of the equipment, nor will it diminish the
capability of the equipment to perform as required during an
accident.
[[Page 1633]]
Combining the maximum per valve leak rate into an overall maximum
leakage limit does not increase the overall permissible leakage and
thus has no significant impact on the consequences of previously
analyzed accidents since the combined leak rate of the main steam
line isolation valves, and thus the contribution of the valves to
overall primary containment leakage as used for analysis purposes,
is unchanged. Extending the drywell nozzle test interval has been
shown by industry experience to not compromise safety, and removing
the specifics of primary containment leakage testing from the
Technical Specifications and referencing 10 CFR Part 50 Appendix J
does not alter either how actual testing is accomplished nor the
acceptance criteria. It has been shown that adopting longer test
intervals based on performance, maintains the safety objective for
containment integrity while at the same time reducing the burden on
licensees, and provides a greater level of worker safety than that
provided by the previous rule.
Therefore, there will be no increase in post accident off-site
or on-site radiation dose as a result of this amendment. The
proposed amendment requires compliance with the regulatory
requirements of 10 CFR Part 50, Appendix J Option B, Section III.A,
for Type A testing that has previously been reviewed by the NRC and
found to be acceptable. Therefore, the amendment will not increase
the consequences of any accident previously evaluated.
b. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed amendment does not involve any modification to
plant equipment or operating procedures, nor will it introduce any
new equipment failure modes that have not been previously
considered. The proposed amendment is limited to changes in
surveillance test frequencies of tests performed while the plant is
in cold shutdown when the associated equipment is not required to be
operable. We therefore conclude the proposed changes will not create
the possibility of a new or different kind of accident from any
accident previously analyzed.
c. The proposed amendment will not involve a significant
reduction in the margin of safety.
Combining the allowable leak rate for the MSIV's from a per
valve limit to an overall limit does not change the total allowable
leakage and therefore post accident dose levels remain unchanged.
Extending the drywell nozzle surveillance test interval from 5 to 10
years has been shown by industry experience to be acceptable.
Extending the intervals between containment integrated leakage tests
as authorized by 10 CFR Part 50, Appendix J, Option B, does not
change the acceptance criteria nor how testing is accomplished.
Based on these considerations, we conclude the proposed
amendment will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: John N. Hannon.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California.
Date of amendment requests: December 19, 1995.
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to relocate Technical Specification (TS)
6.5, ``Review and Audit,'' 6.8, ``Procedures and Programs,'' Sections
6.8.1c., 6.8.1d., 6.8.2, and 6.8.3, in accordance with guidance in an
NRC letter dated October 25, 1993, from William T. Russell to the
chairpersons of industry owners groups and the Commission's Final
Policy Statement on TS Improvements for Nuclear Power Reactors on
relocation of TS that do not satisfy the retention criteria. As part of
the relocation of TS 6.8.2, TS 6.1.1 would be revised to require that
proposed tests, experiments, or modifications that affect nuclear
safety be approved by the plant manager or his designee prior to
implementation.
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 changes simplify the Technical Specifications (TS),
meet regulatory requirements for relocated TS, and implement the
recommendations of: (1) the NRC's letter dated October 25, 1993,
from William T. Russell to the chairpersons of the industry owners
groups; (2) the Commissions's Final Policy Statement on TS
Improvements; and (3) the recently revised 10 CFR 50.36. Future
changes to these requirements will be controlled by 10 CFR 50.54 and
10 CFR 50.59. Any changes that reduce the effectiveness of the
Quality Assurance Program will be approved by the NRC prior to
implementation. The proposed changes are administrative in nature
and do not involve any modifications to any plant equipment or
affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety-related system performs
its function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not alter the basic regulatory
requirements and do not affect any safety analyses. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York.
Date of amendment request: September 15, 1995.
Description of amendment request: The licensee proposes to extend
the surveillance test intervals for the auxiliary electrical systems to
support 24-month operating cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the James A. Fitzpatrick plant in accordance with
the proposed
[[Page 1634]]
Amendment would not involve a significant hazards consideration as
defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes increase the interval between auxiliary
electrical system functional tests and also propose additional
requirements for battery performance testing. These changes are
consistent with the guidance provided in Generic Letter 91-04. These
changes do not involve any special changes to the plant, nor do they
alter the way the auxiliary electrical system functions. Past
equipment performance indicates that the test acceptance criteria
has been consistently met, providing additional assurance that the
longer surveillance interval will not degrade system performance.
The proposed changes revise Bases section 4.9 to clarify battery
testing requirements and indicate consistence with the length of the
24 month operating cycle. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes increase the interval between auxiliary
electrical system functional tests and also propose additional
requirements for battery performance testing. These changes are
consistent with the guidance provided in Generic Letter 91-04. The
proposed changes do not change the ability of the auxiliary
electrical systems to provide electrical power during a design basis
accident. Past equipment performance indicates that the test
acceptance criteria has been consistently met, providing additional
assurance performance. The proposed changes do not modify the design
or operation of plant equipment, therefore, no new or different
failure modes are introduced. The proposed changes revise Basis
section 4.9 to clarify battery testing requirements and indicate
consistency with the length of the 24 month operating cycle.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes increase the interval between auxiliary
electrical system functional tests and also propose additional
requirements for battery performance testing. These changes are
consistent with the guidance provided in Generic Letter 91-09. The
proposed changes do not alter the configuration of the auxiliary
electrical system nor change the manner in which the system
functions. Operation of the facility remains unchanged by the
proposed changes. An evaluation of past equipment performance
indicates that auxiliary electrical system operability is not time
dependent. The proposed changes revise Bases section 4.9 clarify
battery testing requirements and indicate consistency with the
length of the 24 month operating cycle. Therefore, a longer
surveillance test interval for the station batteries and LPCI [low-
pressure coolant injection] batteries will not degrade performance
of the auxiliary electrical system and will not involve a
significant reduction in a margin of safety.
The NRC staff has revised the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York.
Date of amendment request: October 25, 1995.
Description of amendment request: The licensee proposes to extend
the surveillance test intervals for the containment systems to support
24-month operating cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 40.19(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability of
consequences of an accident previously evaluated.
The proposed changes do not involve any physical changes to the
plant, do not alter the way the containment systems function, and
will not degrade the performance of the containment systems. The
type of testing and the corrective actions required if the subject
surveillance fail remains the same. The proposed changes do not
adversely affect the availability of the containment systems or
affect the ability of the systems to meet their design objectives. A
historical review of surveillance test results indicated that there
was no evidence of any failures which would invalidate the above
conclusions.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not modify the design or operation of
the plant and therefore no new failure modes are introduced. No
changes are proposed to the type and method of testing performed,
only to the length of the surveillance interval. Past equipment
performance and on-line testing indicate that longer test intervals
will not degrade the containment systems. A historical review of
surveillance test results indicated that there was no evidence of
any failure which would invalidate the above conclusions.
3. Involve a significant reduction in a margin of safety.
Although the proposed changes will result in an increase in the
interval between surveillance tests, the impact on system
reliability is minimal. This is based on more frequent on-line
testing and the redundant design of the containment systems. A
review of past surveillance history has shown no evidence of failure
which would significantly impact the reliability of the containment
systems. Operation of the plant remains unchanged by the proposed
containment system surveillance test interval extensions. The
assumptions in the Plant Licensing Basis are not impacted. Therefore
the proposed changes do not result in a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York.
Date of amendment request: November 30, 1995.
Description of amendment request: The licensee proposes to extend
the surveillance test intervals for the standby liquid control (SLC)
system to support 24 month operating cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.19(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92 since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
[[Page 1635]]
The proposed changes do not involve any physical changes to the
plant, do not alter any SLC system functions, and will not degrade
the performance of the SLC system. The type of testing and the
corrective actions required if the subject SLC surveillances fail
remain the same. The proposed changes do not adversely affect the
availability of the SLC system or the ability of the system to bring
the reactor from full power to a cold shutdown condition in the
unlikely event that control rods cannot be inserted. A historical
review of SLC surveillance test results indicated that there was no
evidence of any failures that would invalidate the above
conclusions.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not introduce any failure mechanisms of
a different type than those previously evaluated since there are no
physical changes being made to the facility. No changes are proposed
to the type and method of testing performed, only to the length of
the surveillance interval. Past equipment performance and on-line
testing indicate the longer test intervals will not degrade SLC
equipment. A historical review of surveillance test results
indicated that there was no evidence of any failures that would
invalidate the above conclusions.
3. Involve a significant reduction in a margin of safety.
Although the proposed changes will result in an increase in the
interval between surveillance tests, the impact on system
reliability is minimal. This is based on more frequent on-line
testing of major system components and the redundant design of the
SLC system. A review of past SLC surveillance history has shown no
evidence of failures that would significantly impact the reliability
of the SLC system. The longer testing intervals do not significantly
impact the SLC safety margins for SLC normal operation, operation
with inoperable components, or sodium pentaborate solution as
described in the bases of the Technical Specifications. Operation of
the plant remains unchanged by the proposed SLC surveillance
interval extensions. The assumptions in the Plant Licensing Basis
are not impacted. Therefore, the proposed changes do not result in a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York.
Date of amendment request: December 14, 1995.
Description of amendment request: The licensee proposes to
incorporate the inservice testing (IST) requirements of Section XI of
the American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code). The proposed change adds a new surveillance
requirement, 4.0.E, which refers to the requirements of Section XI of
the ASME Code and Addenda established by 10 CFR 50.55a(f). Ancillary
changes are also required since the proposed specification 4.0.E
replaces the surveillance testing requirements of safety related pump
and motor-operated valves and extends the surveillance testing
frequency of other components from once every month, to coincide with
the ASME Code Section XI requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The changes identified in this proposed amendment revise
surveillance testing for various systems based upon the Section XI
of the American Society of Mechanical Engineers [***] Boiler and
Pressure Vessel [***] Code [ASME Code]. None of these changes
involves a hardware modification to the plant, a change to system
operation, a change to the manner in which the system is used, or a
change in the ability of the system to perform its intended
function.
The use of Section XI of the ASME [***] Code as a basis for
establishing surveillance testing and acceptance criteria will not
alter existing accident analyses. This has been acknowledged and
accepted by the NRC in the Standard Technical Specifications. The
change to surveillance testing frequencies reduces testing at power,
increases the availability of systems important to the mitigation of
a DBA [design-basis accident], and minimizes component degradation
due to excessive testing. The ASME [***] Code, Section XI testing
tracks component performance allowing identification of component
degradation and the code specifies that if a pump parameter enters
the alert range, then the testing frequency is doubled until the
cause of the degradation is determined and the condition corrected.
Similarly, if a valve stroke time degrades, the valve testing
frequency is increased to once per month until the cause is
determined and the condition corrected.
The editorial changes are strictly non technical in nature with
no effect on existing analyses. They clarify the Technical
Specifications by improving the legibility of this document.
2. Create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed changes involve no hardware changes, no changes to
the operation of the systems, and do not change the ability of the
systems to perform their intended functions. The use of ASME Section
XI as the basis for testing involves the same testing alignments and
practices previously used as part of either the IST program or
Technical Specification Surveillance Requirements. The editorial
changes have no effect on plant practices.
3. Involve a significant reduction in the margin of safety.
There are no hardware modifications, changes to system
operations, or effect on the ability of systems to perform their
intended function associated with the proposed changes. The proposed
changes to reference pump and valve testing to Section XI of the
ASME [***] Code and remove individual Surveillance Requirements in
the Technical Specifications does not relax any controls or
limitations. The resulting reduction in test frequency, while
reducing the possibility of detecting a degraded component prior to
failure, is offset by the increased availability of systems
important to plant safety and an associated reduction in component
wear and degradation due to excessive testing. Additionally, the
ASME testing program evaluates components for degraded performance
and will identify such degradation early. There are no safety
margins associated with the editorial corrections.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina.
Date of amendment request: December 8, 1995.
Description of amendment request: The proposed changes add a new
[[Page 1636]]
surveillance requirement to Technical Specification (TS) Section
4.1.2.2 and deletes TS Sections 3/4.1.2.3 and 3/4.1.2.4 associated with
the Borations Systems section. TS Section 3/4.9.3 is being revised to
assure only one charging pump is capable of Reactor Coolant System
injection in the applicable modes and to add a new surveillance
requirement to demonstrate this assurance. TS Section 4.5.2.f is being
revised to delete specific Emergency Core Cooling System pump testing
acceptance criteria and reference acceptance criteria located in the
plant Inservice Testing Program. In addition, the licensee has proposed
changes to the bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability or consequences of an accident previously
evaluated is not significantly increased.
The implementation of the above described TS changes will have
no impact on the probability of an accident occurring. The testing
of the ECCS pumps at a more appropriate point on their
characteristic curve is not a precursor to an accident. There is no
hardware, software, or testing methodology change proposed that
would decrease confidence in the reliability of these systems/
components.
The proposed revision to the ECCS Pump testing surveillance will
allow greater flexibility for testing and will provide more useful
information about the performance capabilities of those pumps.
The deletion of the Reactivity Control System Specifications
(Charging Pumps - Operating and Charging Pumps - Shutdown) will have
no impact on the capability of the Charging/SI pumps to perform
their design function. The additional Action Statement and
Surveillance for low temperature overpressure (LTOP) assure that
safety analyses remain valid and initial conditions are not changed.
The additional Surveillance Requirement for Boration Systems assures
that one charging pump will be operable during Modes 5 and 6.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
This proposed TS change does not involve any changes to station
hardware, software, or operating practices. The changes do provide
for a revision to the testing methodology used in demonstrating the
capability of the ECCS pumps.
This methodology will test the ECCS pumps at a point on the
pump's characteristic curve that will more reliably indicate the
pump's continued operability at or near the parameters the pump
would be required to provide during a postulated accident.
The deletion of the Reactivity Control System Specifications
(Charging Pumps - Operating and Charging Pump - Shutdown) will not
provide additional challenges to the capability of the plant to meet
normal operational needs or mitigate the conditions of a design
basis accident. The ECCS Subsystems TS provide similar surveillance
requirements to insure continued operability of the Charging/SI
pumps. The LTOP TS will now provide requirements to assure that
design assumptions are not challenged and RCS integrity is
maintained.
Therefore, as the above described change has no impact on plant
performance, the possibility of a new or different kind of accident
being created as a result of this change is negligible.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
The change in testing philosophy for ECCS pumps should bring an
increase in margin of safety, since testing will be conducted at
reference flow points closer to actual pump parameters for accident
conditions. For the Residual Heat Removal Pumps this will be
conducted quarterly and for the centrifugal charging pumps, they
will be tested quarterly on minimum flow and each refueling outage
at substantial flow per the Inservice Testing Program.
The surveillance requirements of TS 3/4.1.2.3 and TS 3/4.1.2.4
are essentially the same as those in 3./4.5.2 and 3/4.5.3 (ECCS
Subsystems), and the deletion of these requirements will have no
adverse impact on margin on safety. The addition of the Action
Statement and Surveillance Requirements to 3/4.4.9.3 (Overpressure
Protective Systems) provide additional requirements to supplement
those above to assure RCS integrity is maintained for all
operational modes. The addition of the Surveillance Requirement to
3/4.1.2.1 will provide assurance that reactivity control can be
maintained for Modes 5 and 6 through the charging system flow path.
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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Project Director: Frederick J. Hebdon.
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: December 19, 1995.
Description of amendments request: The proposed amendments would
replace the requirements associated with the Control Room Emergency
Ventilation System with requirements related to the operation of the
Control Room Emergency Filtration/Pressurization System and Control
Room Air Conditioning System. These changes are technically consistent
with the requirements of NUREG-1431, Revision 1, ``Westinghouse
Standard Technical Specifications,'' issued on April 7, 1995. Also, a
one-time extension to the allowable outage time for the control room
recirculation filtration system is included to facilitate
implementation of design modifications to enhance the reliability of
the control room air conditioning system during the spring of 1996.
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:
Based on the preceding evaluation, the following conclusions are
provided with respect to the criteria contained in 10 CFR 50.92.
(1) The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the FSAR [Final Safety Analysis Report]. The proposed changes have
no impact on the probability of an accident. The control room
ventilation systems are support systems which have a role in the
detection and mitigation of accidents but do not contribute to the
initiation of any accident previously evaluated. Reorganizing the
technical specifications by functions have no impact on the course
of any accidents previously evaluated. The other changes which are
being made improve the ability to mitigate fuel handling accidents.
Specifying an allowed outage time (AOT) of 30 days for the cooling
of recirculated air while one train is inoperable is based on the
significance of the cooling function but does represent an increase
in the allowed outage time and thus an increase in the probability
that the functions could be unavailable. This increase is not
considered significant based on several factors including: the
design is based on the worst postulated meteorological conditions;
generally, less than design cooling is required and a partial
failure in the system may have no impact; and unavailability failure
does not create an immediate irreversible impact (i.e., temperature
will increase slowly over a period of time); the system could be
restored or its loss mitigated without any impact on the course or
whatever accident is being considered; and the extended AOT would
allow more opportunity to perform major required maintenance and
thus may provide an overall improvement in equipment reliability.
In addition, the one-time change to the AOT for the
recirculation filtration will not
[[Page 1637]]
significantly increase the probability or consequences of an accident
due to the low probability of an event result[ing] in an airborne
release of radioactivity. Such an event requires multiple failures
of safety systems that are governed by technical specifications not
affected by these changes. In addition, compensatory measures have
been identified that limit the potential exposure of control room
operators in response to a postulated release.
The net effect of these changes is not significant and, as a
result, the changes do not involve a significant increase in the
consequences of an accident previously evaluated.
(2) The proposed changes to the Technical Specifications do not
increase the possibility of a new or different kind of accident than
any accident already evaluated in the FSAR. No new limiting single
failure or accident scenarios have been created or identified due to
the proposed changes. Safety-related systems are expected to perform
as designed. Although the changes could have a minor impact on the
air conditioning system availability, the changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) The proposed changes do not involve a significant reduction
in the margin of safety. The changes proposed do not alter the
environmental conditions which are to be maintained in the control
room during normal operations and following an accident. As a
result, the margin of safety for these functions remains the same.
Although there is a potential impact on the air conditioning
system's postulated availability, there is no impact on the accident
analyses. Further, although the one-time AOT extension for the
recirculation filtration system increases the system unavailability
during the planned CRACS [Control Room Air Conditioning System]
design changes, the net effect is a benefit to plant safety due to
the enhancement to control room cooling capability. Thus, even if
system availability issues were considered an aspect of margin of
safety, 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: 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.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama.
Date of amendment request: December 8, 1995 (TS 364).
Description of amendment request: The licensee proposes revision of
Units 1, 2, and 3 Technical Specifications (TS) Section 4.7.A to
implement the revision to 10 CFR 50, Appendix J. The new rule (Option
B) provides a voluntary performance-based testing option for
containment leak rate testing. Option B containment leak rate testing
requirements are based on system and component performance in lieu of
compliance with the current prescriptive requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment to TS Section 4.7.A is in accordance with
Option B to 10 CFR 50, Appendix J. The proposed amendment adds a
voluntary performance based option for containment leak rate
testing. The changes being proposed do not affect the precursor for
any accident or transient analyzed in Chapter 14 of the BFN [Browns
Ferry Nuclear Plant] Updated Final Safety Analysis Report (UFSAR).
The proposed change does not increase the total allowable primary
containment leakage rate. The proposed change does not reflect a
revision to the physical design and/or operation of the plant.
Therefore, operation of the facility in accordance with the proposed
change does not affect the probability or consequences of an
accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment to TS Section 4.7.A is in accordance with
the new performance-based option (Option B) to 10 CFR 50, Appendix
J. The changes being proposed will not change the physical plant or
the modes of operation defined in the facility license. The proposed
changes do not increase the total allowable primary containment
leakage rate. The changes do not involve the addition or
modification of equipment, nor do they alter the design or operation
of plant systems. Therefore, operation of the facility in accordance
with the proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to TS Section 4.7.A is in accordance with
the new option to 10 CFR 50, Appendix J. The proposed option is
formulated to adopt performance-based approaches. This option
removes the current prescriptive details from the TS. The proposed
changes do not affect plant safety analyses or change the physical
design or operation of the plant. The proposed change does not
increase the total allowable primary containment leakage rate.
Therefore, operation of the facility in accordance with the proposed
change does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio.
Date of amendment request: December 12, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.6.1.1, Containment Systems--
Primary Containment--Containment Integrity; TS 3/4.6.1.2, Containment
Systems--Containment Leakage; TS 3/4.6.1.6, Containment Systems--
Containment Vessel Structural Integrity; TS 3/4.6.5.3, Containment
Systems--Shield Building Structural Integrity; and associated Bases.
The proposed revisions adopt the provisions of Appendix J, Option B for
Type A containment leakage testing as modified by approved exemptions
and in accordance with the guidance of Regulatory Guide 1.163. The
licensee proposes to delete surveillance requirement (SR) 4.6.1.2, SR
4.6.1.2.b, SR 4.6.1.2.c, and SR 4.6.1.2.i since these requirements
contain details that are now included in standards that are referenced
by Regulatory Guide 1.163. TS 3/4.6.1.6 and TS 3/4.6.5.3 which address
containment building and shield building structural integrity are
proposed to be deleted since the requirements are addressed in revised
TS 3.6.1.2.a. The licensee proposes to delete the exemption included in
Bases
[[Page 1638]]
3/4.6.1.2 since it is no longer applicable. Additionally, the licensee
proposes to modify the Action statement associated with TS 3.6.1.2 to
reflect the action to take if the as-left rather than the as-found
leakage exceeds 0.75 La.
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:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in
accordance with the changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because accident initiators,
conditions, or assumptions are not affected by the proposed changes.
The proposed changes to the Technical Specifications implement
10 CFR 50 Appendix J Option B for Type A testing, including visual
examinations of the containment vessel and shield building, and make
various administrative changes to the Technical Specifications and
associated Technical Specification Bases. Therefore, as stated
above, these proposed changes do not affect accident initiators,
conditions, or assumptions.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
change the source term, containment isolation, or allowable
releases.
The proposed changes involve containment leakage testing and
test frequency. The allowable containment leakage rates presently
specified in the Technical Specifications remain unchanged.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes.
3. Not involve a significant reduction in a margin of safety,
for the reasons cited below.
The proposed changes involve containment leakage testing and
test frequency. The allowable containment leakage rates presently
specified in the Technical Specifications remain unchanged. The
Technical Specifications, under the proposed changes, will continue
to ensure containment system reliability by periodic testing
performed in full compliance with 10 CFR 50 Appendix J.
As stated in the Federal Register publication of the final rule,
60 FR 49495 dated September 26, 1995, the final rule improves the
focus of the regulations by eliminating prescriptive requirements
that are marginal to safety. Further, the final rule allows test
intervals to be based on system and component performance and
provides licensees greater flexibility for cost-effective
implementation methods of regulatory safety objectives. The final
rule publication also discusses the following specific findings
documented in NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' September, 1995, which justify the proposed change in
frequency of Type A Integrated Leak Rate Testing (ILRT):
1. The fraction of leakages detected only by ILRT's is small, on
the order of a few percent.
2. Reducing the frequency of ILRT testing from 3 every 10 years
to one every 10 years leads to a marginal increase in risk.
3. At a frequency of one test every 10 years, industry-wide
occupational exposure would be reduced by 0.087 person-sievert (8.7
person-rem) per year.
Based on these considerations, it is concluded that the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin.
Date of amendment request: December 13, 1995.
Description of amendment request: The proposed amendments will
modify Technical Specification (TS) Sections 15.1, ``Definitions,''
15.2, ``Safety Limits and Limiting Safety System Settings,'' 15.3,
``Limiting Conditions for Operation,'' and 15.6, ``Administrative
Controls.'' The proposed changes would modify the TSs to account for
the creation and maintenance of a Core Operating Limits Report.
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 this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
The relocation of the cycle-specific parameters from the Point
Beach Nuclear Plant (PBNP) Technical Specifications to the Core
Operating Limits Report (COLR) has no impact on plant operation or
accident analyses. The proposed changes are administrative in
nature. The Technical Specifications will continue to require
operation within the core operational limits for each cycle reload
calculated by the NRC-approved reload design methodologies. The
appropriate actions required if limits are exceeded will remain in
the Technical Specifications. The reload report presents the results
of a cycle-specific evaluation of accidents and transients addressed
in the PBNP Final Safety Analysis Report (FSAR). The cycle-specific
evaluation demonstrates that changes in the unit's fuel cycle design
and corresponding COLR parameters do not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Therefore, these changes do not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. Operation of this facility under the proposed Technical
Specifications will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change to relocate the cycle-specific parameters
from the Technical Specifications to the COLR is administrative in
nature. No change to the design, configuration, or method of
operation of the plant is made by this change. The cycle-specific
parameters will be determined using NRC-approved methodologies. The
Technical Specifications will continue to require operation within
the core operating limits and appropriate actions will be taken if
the limits are exceeded.
Therefore, these changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Operation of this facility under the proposed Technical
Specifications will not create a significant reduction in a margin
of safety.
Existing Technical Specification operability and surveillance
requirements are not reduced by the proposed changes to relocate
cycle-specific parameters from the Technical Specifications to the
COLR. The cycle-specific COLR limits for reloads will continue to be
developed based on NRC-approved methodologies, thereby maintaining
accepted margins of safety. The Technical Specifications will still
require that the core be operated within these limits and specify
appropriate actions to be taken if the limits are violated. Each
reload undergoes a 10 CFR 50.59 safety review to assure that
operating the unit within the cycle-specific limits will not involve
a significant reduction in a margin of safety. Therefore, these
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: Joseph P. Mann Library, 1516
[[Page 1639]]
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas.
Date of amendment request: December 13, 1995.
Description of amendment request: This license amendment request
proposes to revise the 125-volt D.C. Sources Technical Specifications
(3.8.2.1 and 3.8.2.2) to include provisions for installed spare
chargers, which will be added to the plant design during the next
refueling outage. The Onsite Power Distribution Technical
Specifications 3.8.3.1 and 3.8.3.2 would be revised to indicate that
spare chargers may be connected in place of the primary chargers.
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.
These proposed technical specification changes do not alter the
plant design bases nor do they involve any hardware changes that
significantly increase the probability of any event initiators.
There will be no change to normal plant operating parameters or
accident mitigation capabilities. There will be no increase in the
consequences of any accident or equipment malfunction.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed technical specification changes do not involve any
design bases changes nor are there any changes to the method by
which any safety-related plant system performs its safety function.
The normal manner of plant operation is unaffected. No new accident
scenarios, transient precursors, failure mechanisms, or limiting
single failures are introduced as a result of these changes.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined, nor will there be
any effect in those plant systems necessary to assure the
accomplishment of protection functions. There will be no impact on
DNBR [departure from nucleate boiling ratio] limits, FQ, F-
delta-H, LOCA [loss-of-coolant accident] PCT [peak cladding
temperature], peak local power density or any other 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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas.
Date of amendment request: December 13, 1995.
Description of amendment request: This change request proposes
revising the minimum and maximum flow requirements for the centrifugal
charging pumps (CCPs) and safety injection pumps (SIPs) specified in
Technical Specification Surveillance Requirement 4.5.2.h. Specifically,
the proposed changes would:
(1) Decrease the minimum limits on the sum of the injection line
flow rates, excluding the highest flow rate, from 346 gpm to 330 gpm
for the CCPs and from 459 gpm to 450 gpm for the SIPs.
(2) Revise the maximum pump flow rate for the SIP from 665 to 670
gpm, but retain the CCPs maximum pump flow rate at its current value of
556 gpm.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, 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 will not result in a condition where the
material or construction standards applicable prior to the change
are altered. The ECCS [emergency core cooling system] system
integrity is not affected by this change, and this change will not
affect the ability of the ECCS to fulfill its design functions. This
change will modify the pump surveillance criteria to prevent pump
runout during the test, but will not affect the method of operation
of the system and will not alter the testing method for the pumps.
This change will slightly alter the acceptance criteria of the test,
but the changes have been determined to be enveloped by the ECCS
pump flow and balance criteria assumed in the safety analyses
described in the USAR [Updated Safety Analysis Report]. This change
will not affect the ability of the ECCS to mitigate the consequences
of any previously evaluated accident. The proposed change will not
alter, degrade or prevent the response of the ECCS to any accident
scenarios evaluated in the USAR. Therefore, neither the probability
of occurrence nor the consequences of any accident previously
evaluated in the USAR will be increased by this change.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will alter the existing ECCS pump flow test
to prevent pump runout during the test by slightly altering the
acceptance criteria of the test. However, the proposed changes have
been determined to be enveloped by the ECCS pump flow and balance
criteria assumed in the safety analyses described in the USAR. This
change will not create a new type of accident or malfunction, and
the method and manner of plant operation remains unchanged. This
change will not alter the safety functions of the ECCS. The safety
design bases in the USAR have not been altered, and no new or
different accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this change. Therefore, the possibility of a new or
different kind of accident other than those already evaluated will
not be created by this change.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There are no changes being made to any safety limits or safety
system settings that would adversely impact plant safety. This
proposed change will have no affect on the availability, operability
or performance of any safety-related system or component. The
analysis results and conclusions of the accidents presented in the
current USAR would not be adversely affected by the revised
surveillance requirements for the ECCS. This conclusion is drawn
based on the evaluation that confirms that the actual ECCS flow
characteristics remain consistent with assumptions used in the WCGS
[Wolf Creek Generating Station] accident analyses. Specifically, the
accident analyses which are limiting with minimized ECCS flow have
already been analyzed using revised ECCS flows that were developed
based on a more conservative minimum flow than the proposed minimum
ECCS flow requirement. For the analyses which are limiting with a
higher ECCS flow, the evaluation indicated that a higher pump runout
limit proposed for the SIPs would have insignificant effect on the
results and conclusions of the analyses. The evaluation also
indicated that the ECCS pump operability would not be a concern as a
result of increasing the SIPs runout limit because the available
runout margin is sufficient to accommodate the cumulative effect of
the ECCS performance issues. Based on these reasons, it is concluded
that
[[Page 1640]]
implementation of the proposed changes will have no adverse impact on
the ECCS subsystems' operability and their intended safety function.
Therefore, the proposed change would not result in a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas.
Date of amendment request: December 13, 1995.
Description of amendment request: This license amendment request
proposes revising Surveillance Requirement 4.1.3.1.3 to delete the
requirement for performing the control rod drop surveillance test with
Tavg greater than or equal to 551 deg.F. This would allow
performing this test with Tavg below 551 deg.F. This change will
also add justification for performing the rod drop test with Tavg
below 551 deg.F to Bases Section 3/4.1.3, ``Movable Control
Assemblies.''
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 will not result in a condition where the
material or construction standards applicable prior to the change
are altered. The rod control system integrity is not affected by
this change, and this change will not affect the ability of the
system to fulfill its design function. This change will allow the
control rod drop test to be performed at lower temperatures than
currently allowed, but will not affect the method of operation of
the system and will not alter the drop time criterion of the test.
This change will not affect any fission product barrier, and will
not affect the integrity of any fuel assembly or the reactor
internals. Thus this change will not affect the ability of the rod
control system to mitigate the consequences of any previously
evaluated accident. The proposed change will not alter, degrade or
prevent the response of the rod control system to any accident
scenarios evaluated in the USAR [Updated Safety Analysis Report].
Therefore, neither the probability of occurrence nor the
consequences of any accident previously evaluated in the USAR will
be increased by this change.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will alter the existing rod drop test to
allow the test to be performed over a range of temperatures, but
will not alter the rod drop time criterion of the test. This change
will not create a new type of accident or malfunction, and the
method and manner of plant operation remains unchanged. This change
will not alter the safety functions of the rod control system. The
safety design bases in the USAR have not been altered, and no new or
different accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this change. Therefore, the possibility of a new or
different kind of accident other than those already evaluated will
not be created by this change.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There are no changes being made to any safety limits or safety
system settings that would adversely impact plant safety. This
proposed change will have no affect on the availability, operability
or performance of any safety-related system or component. The change
will not prevent inspections or surveillances required by the
technical specifications, and does not alter the rod drop time
criterion specified in the technical specifications. Performance of
the rod drop tests at other temperatures allows an alternative
method to verify that the rod drop time currently specified in the
technical specifications and used in the safety analyses continues
to be valid. Therefore, the proposed change would not result in a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
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.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania.
Date of amendment request: November 21, 1995
Brief description of amendment request: The proposed amendments
would revise surveillance requirements for the high pressure coolant
injection and reactor core isolation cooling systems and would make an
administrative change to Section 5.5.7 of the technical specifications
to eliminate reference to a section which was previously eliminated.
Date of publication of individual notice in Federal Register:
December 5, 1995 (60 FR 62271).
Expiration date of individual notice: January 3, 1996.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania.
Date of amendment request: November 30, 1995.
[[Page 1641]]
Brief description of amendment request: The proposed amendments
would revise the minimum allowable control rod scram accumulator
pressure and charging water header pressure from a value of 955 psig to
a value of 940 psig.
Date of publication of individual notice in Federal Register:
December 8, 1995 (60 FR 63073).
Expiration date of individual notice: January 8, 1996.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania.
Date of amendment request: December 19, 1995.
Brief description of amendment request: The proposed amendment
would revise the ventilation filter test program (VFTP) bypass and
penetration leakage test acceptance criteria from less than 0.05
percent to less than 1.0 percent. The change corrects an administrative
error that occurred during the development of the Peach Bottom Improved
Technical Specifications which were issued as Amendments 210 and 214 to
the Peach Bottom licenses on August 30, 1995.
Date of publication of individual notice in Federal Register:
December 27, 1995 (60 FR 66997).
Expiration date of individual notice: January 25, 1996.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
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.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland.
Date of application for amendment: October 20, 1995.
Brief description of amendment: The one-time amendment revises the
Calvert Cliffs Nuclear Power Plant, Unit No. 1 Technical Specifications
by extending certain 18-month instrument surveillance intervals by a
maximum of 39 days to March 31, 1996. This amendment will be superseded
by Amendment No. 208 when it is implemented prior to restart from the
Unit No. 1 spring 1996 refueling outage.
Date of issuance: December 28, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 209.
Facility Operating License No. DPR-53: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58396).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland.
Date of application for amendment: October 2, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications regarding allowable outage time (AOT) associated with
the control room emergency ventilation system. It extends the AOT for
one train from 7 days to 30 days on a one-time basis (for the loss of
the emergency power supply only) to allow for modifications during the
upcoming Unit No. 1 refueling outage in the spring of 1996.
Date of issuance: December 19, 1995.
Effective date: As of the date of issuance to be implemented during
the Unit No. 1 spring 1996 refueling outage.
Amendment No.: 187.
Facility Operating License No. DPR-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56363).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 19, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
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 amendments: September 10, 1993, as
supplemented on June 16, 1995.
Brief description of amendments: This application upgrades the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Section 3/4.8 (Plant Systems).
Date of issuance: December 19, 1995.
[[Page 1642]]
Effective date: Immediately, to be implemented no later than June
30, 1996.
Amendment Nos.: 144, 138, 166, and 162.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37086).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 19, 1995.
No significant hazards consideration comments received: No.
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.
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 amendments: September 15, 1995.
Brief description of amendments: The amendments upgrade the current
custom Technical Specifications (TS) for Dresden and Quad Cities to the
Standard Technical Specifications contained in NUREG-0123, ``Standard
Technical Specification General Electric Plants BWR/4.'' The
application dated September 15, 1995, contains some of the TSUP open
items from previous Dresden and Quad Cities TS amendments issued by the
NRC.
Date of issuance: December 19, 1995.
Effective date: Immediately, to be implemented no later than June
30, 1996.
Amendment Nos.: 145, 139, 167 and 163
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 5, 1995 (60 FR
52220).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 19, 1995.
No significant hazards consideration comments received: No.
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.
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 amendments: September 17, 1993, as
supplemented July 28, 1995.
Brief description of amendments: This application upgrades the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Section 3/4.5 (Emergency Core
Cooling Systems).
Date of issuance: December 27, 1995.
Effective date: Immediately, to be implemented no later than June
30, 1996.
Amendment Nos.: 146, 140, 168, and 164.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42599).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 27, 1995.
No significant hazards consideration comments received: No.
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.
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: November 14, 1995.
Brief description of amendments: These amendments change the
implementation dates of all previous TSUP amendments from December 31,
1995, to no later than June 30, 1996.
Date of issuance: December 29, 1995.
Effective date: December 29, 1995.
Amendment Nos.: 147 and 141.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the license.
Date of initial notice in Federal Register: November 29, 1995 (60
FR 61272).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 29, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station,
Unit 1, LaSalle County, Illinois.
Date of application for amendment: October 2, 1995.
Brief description of amendment: The amendment revises the safety/
relief valve (SRV) safety function lift setting allowable tolerance
band from -3/+1% to 3% and includes a requirement for the
lift settings to be within 1% of the technical
specification limit following testing.
Date of issuance: January 3, 1996.
Effective date: Upon date of issuance; shall be implemented prior
to the restart of Unit 1 from its seventh refueling outage.
Amendment No.: 108.
Facility Operating License No. NPF-11: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58398).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 3, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
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: September 5, 1995.
Brief description of amendments: In Section 5.2.5 of the Catawba
Safety Evaluation Report (SER, NUREG-0954), the NRC staff identified
that the air particulate monitors (EMF38, at both Units 1 and 2), are
designed to seismic Category I requirements. A recent engineering
review by the licensee determined that documentation did not exist to
show these monitors are designed to seismic Category I requirements. In
a submittal dated September 8, 1994, the licensee proposed a technical
justification for not requiring the subject monitors to be
[[Page 1643]]
seismic Category I, and by letter dated September 5, 1995, provided
additional justification and requested amendments to the licenses for
both Units 1 and 2. The NRC staff has reviewed the licensee's
justification and concludes that the containment air particulate
monitors at Catawba do not have to meet seismic Category I
requirements. The bases for this conclusion are included in the NRC
staff's Safety Evaluation.
Date of issuance: December 29, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--140; Unit 2--134.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: November 28, 1995 (60
FR 58690).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 29, 1995 and an Environmental
Assessment dated December 22, 1995.
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. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina.
Date of application for amendments: September 1, 1995, as
supplemented by letters dated October 17 and November 15, 1995.
Brief description of amendments: The requested changes would revise
Technical Specification (TS) 6.9.1.9 to include references to updated
or recently approved methodologies used to calculate cycle-specific
limits contained in the Core Operating Limits Report (COLR). The
subject references have previously been reviewed and approved by the
NRC staff.
Date of issuance: December 19, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--160; Unit 2--142.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 25, 1995 (60 FR
54718).
The October 17 and November 15, 1995, letters provided clarifying
information that did not change the scope of the September 1, 1995,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 19, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina.
Date of application for amendments: January 12, 1995, as
supplemented by letter dated June 29, 1995.
Brief description of amendments: The amendments would revise and
clarify portions of Technical Specification Section 6.0,
``Administrative Controls.''
Date of issuance: December 19, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--161; Unit 2--143.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14018).
The June 29, 1995, letter provided clarifying information that did
not change the scope of the January 12, 1995, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated December 19, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
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: July 26, 1995, as supplemented
by letter dated November 20, 1995.
Brief description of amendments: The amendments add a footnote to
Technical Specification 3.7.8 to provide for a one-time extension of
the allowable outage time from 72 hours to 7 days for the Oconee
overhead emergency power path to be inoperable, so that proposed
modifications to the degraded grid protection system and the external
grid trouble protection system may be performed.
Date of Issuance: December 27, 1995.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1--213; Unit 2--213; Unit 3--210.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42601).
The November 20, 1995, letter provided clarifying information that
did not change the scope of the July 26, 1995, application and the
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated December 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas.
Date of application for amendment: July 19, 1995.
Brief description of amendment: The amendment reduced the
requirements associated with the exercise frequency of control element
assemblies from once per 31 days to once per 92 days.
Date of issuance: December 22, 1995.
Effective date: December 22, 1995, to be implemented within 30
days.
Amendment No.: 173.
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52929).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas.
Date of application for amendment: April 4, 1995.
Brief description of amendment: The amendment revises surveillance
[[Page 1644]]
requirements associated with the main turbine steam valves.
Date of issuance: December 22, 1995.
Effective date: December 22, 1995, to be implemented within 30
days.
Amendment No.: 174.
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35069).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida.
Date of application for amendments: September 11, 1995, as
supplemented by letter dated November 22, 1995.
Brief description of amendments: These amendments revise the
emergency diesel generator testing requirements to incorporate the
recommendations of Generic Letters 93-05 and 94-01.
Date of issuance: December 28, 1995.
Effective date: December 28, 1995.
Amendment Nos. 181 and 175.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52930).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,
Georgia.
Date of application for amendments: December 2, 1994.
Brief description of amendments: The amendments replace Appendix B,
``Environmental Technical Specifications,'' with an Environmental
Protection Plan (Nonradiological) and revise the Operating Licenses to
reflect these changes.
Date of issuance: December 19, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--199; Unit 2--140.
Facility Operating License Nos. DPR-57 and NPF-5. Amendments
revised the Technical Specifications and Operating Licenses.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
502).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 19, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook,
Nuclear Plant, Unit No. 1, Berrien County, Michigan.
Date of application for amendment: April 13, 1995, as supplemented
August 28 and October 27, 1995.
Brief description of amendment: The amendment modifies the
Technical Specifications to allow use of laser-welded sleeves to repair
defective steam generator tubes.
Date of issuance: January 4, 1996.
Effective date: January 4, 1996, with full implementation within 45
days.
Amendment No.: 205.
Facility Operating License No. DPR-58. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29877).
The August 28 and October 27, 1995, supplements provided clarifying
information and updated Technical Specification pages. These
supplements did not change the proposed no significant hazards
considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 4, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear
Power Station, Unit 1, New London County, Connecticut.
Date of application for amendment: August 31, 1995, as supplemented
December 5, 1995.
Brief description of amendment: The amendment modifies the
definition of HOT SHUTDOWN and COLD SHUTDOWN to specify that the
definitions are not applicable during the performance of an inservice
hydrostatic and leak test (IHLT). Technical Specification Section 3.6.B
and 4.6.B is modified by adding Section 3.6.B.1.b and 4.6.B.1.b to
identify the requirements that must be satisfied to consider the
reactor in COLD SHUTDOWN during the performance of an IHLT. In
addition, the amendment changes temperature specific requirements on
several pages to mode or condition specific requirements; makes several
editorial changes; and changes the associated Bases.
Date of issuance: December 29, 1995.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 90.
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49940).
The December 5, 1995, 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 December 29, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut.
Date of application for amendment: May 1, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to extend the interval for performance of selected
surveillances to accommodate a 24-month fuel cycle. Specifically, this
amendment changes the definition for a refueling interval, changes the
BASES for surveillances that are performed at least once each fuel
cycle and changes the surveillance frequencies for:
(1) The flow path tests of the boron injection system,
(2) The operability tests of the digital rod position indicatiors,
(3) The drop time of the full-length shutdown and control rods,
(4) The channel calibration of the loose-part detection system,
[[Page 1645]]
(5) The channel calibration of the seismic monitoring
instrumentation,
(6) The activation of the pumps and the flow path tests of the
valves in the containment quench and recirculation spray systems and
(7) The tests of the intended actuation positions of the
containment isolation valves.
Date of issuance: December 28, 1995.
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendment No.: 122.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58402).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut.
Date of application for amendment: July 17, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications pertaining to the plant air filtration and ventilation
systems to extend the surveillance frequencies that are now required to
be performed at least once per 18 months to specify that the
surveillances are to be performed at least once each refueling
interval.
Date of issuance: December 28, 1995.
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendment No.: 123.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58402).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut.
Date of application for amendment: July 14, 1995.
Brief description of amendment: The amendment revises the frequency
of those surveillance requirements for the emergency core cooling
systems that now require that the surveillances be performed ``at least
once per 18 months'' to specify that the surveillances be performed
``at least once each refueling interval.''
Date of issuance: December 28, 1995.
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendment No.: 124.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58402).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California.
Date of application for amendments: September 29, 1995.
Brief description of amendments: The amendments added a one-time
footnote to the Technical Specifications related to the diesel
generator fuel oil storage and transfer system to permit each of the
existing storage tanks to be removed from service for up to 60 days so
they can be replaced with double walled tanks and piping that comply
with new California regulations.
Date of issuance: January 3, 1996.
Effective date: January 3, 1996, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 1--Amendment No. 109; Unit 2--Amendment No.
108.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58403).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 3, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California.
Date of application for amendment: October 8, 1993, as supplemented
October 28, 1994.
Brief description of amendment: This amendment revised the
Technical Specification by deleting Figure II-2, ``Restricted Area Per
10 CFR 20.3(a)(14)'' and by deleting the restricted area boundary line
from Figure V-3, ``HBPP Groundwater Monitoring System Wells.''
Date of issuance: December 21, 1995.
Effective date: This license amendment is effective as of the date
of its issuance and must be fully implemented no later than 30 days
from the date of issuance.
Amendment No.: 30.
Facility License No. DPR-7: This amendment revised the TS.
Date of initial notice in Federal Register: January 5, 1994 (59 FR
624).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Humboldt County Library, 1313
3rd Street, Eureka, California 95501.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania.
Date of application for amendments: March 31, 1995.
Brief description of amendments: The amendments incorporate a
change in the Station Technical Specifications for both units that
modifies the requirement in TS 4.4.4.3.a to have the pH of the reactor
coolant measured every 72 hours. The amendments add the clarification
that the pH measurement will be performed only when the coolant
conductivity is greater than 1.0 micro-mho/cm at 25 deg.C ( deg.77).
Date of issuance: January 3, 1996.
[[Page 1646]]
Effective date: Both units, as of date of issuance and are to be
implemented within 30 days.
Amendment Nos.: 156 and 127.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20522).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 3, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Pennsylvania Power and Light Company, Docket No. 50-388,
Susquehanna Steam Electric Station, Unit 2, Luzerne County,
Pennsylvania.
Date of application for amendment: August 11, 1995.
Brief description of amendment: The amendment revises the Unit 2
Technical Specifications (TSs) to reestablish the original operability
requirements for the Neutron Flux function, and to delete the footnote
that was added to TS page 3/4 3-71 under Amendment No. 115, regarding
the length of time that the revised operability values were valid.
Date of issuance: January 3, 1996.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment No.: 128.
Facility Operating License No. NPF-22. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47623).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 3, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York.
Date of application for amendment: May 12, 1995.
Brief description of amendment: The amendment modifies the
Technical Specifications (TSs) to extend the surveillance test
intervals for the emergency service water system to support 24-month
operating cycles. Surveillance test interval extensions are denoted as
being performed ``every 24 months'' or ``at least once per 24 months''
consistent with the guidance provided in Generic Letter (GL) 91-04,
``Changes in Technical Specification Surveillance Intervals to
Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991. The NRC staff
has determined that the proposed TS changes are in accordance with GL
91-04, and are therefore acceptable.
Date of issuance: December 21, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 230.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47623)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina.
Date of application for amendment: February 21, 1995, as
supplemented on August 31, 1995, and December 4, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) support of the licensee's plan to implement the
revised 10 CFR Part 20, ``Standards for Protection Against Radiation.''
Also, several editorial changes to improve the clarity of the TS were
made.
Date of issuance: December 28, 1995.
Effective date: 90 days after issuance.
Amendment No.: 130.
Facility Operating License No. NPF-12. Amendment revises the
operating license.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16200). Renoticed on September 27, 1995 (60 FR 49946) due to changes in
the licensee's proposed no significant hazards consideration analysis
that were included in the August 31, 1995 supplemental letter. The
December 4, 1995 letter provided supplemental information that did not
change the second proposed no significant hazards consideration. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated December 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of application for amendment: June 21, 1994, as supplemented
by letter dated October 23, 1995.
Brief description of amendment: The amendment revises Technical
Specification (TS) 6.5.1, 6.5.2 and 6.5.3 to relocate the review and
audit requirements of the On-site Review Committee (ORC) and the
Nuclear Safety Review Board (NSRB) to the Operational Quality Assurance
Manual (OQAM). In addition, the amendment deletes reference to the
Manager, Nuclear Safety and Emergency Preparedness, in TS 6.2.3. The
Index is revised to reflect the relocations.
Date of issuance: December 26, 1995.
Effective date: December 26, 1995, to be implemented within 30 days
from the date of issuance.
Amendment No.: 107.
Facility Operating License No. NPF-30. The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45036) and November 27, 1995 (60 FR 58406). The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
December 26, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: July 20, 1995.
Brief description of amendments: These amendments establish a new
setpoint for the steam generator high-high level and provide more
restrictive setting limits for certain reactor protection system/
engineered safety features actuation system setpoints.
Date of issuance: December 28, 1995.
Effective date: December 28, 1995.
Amendment Nos. 206 and 206.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45190).
[[Page 1647]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin.
Date of application for amendment: September 19, 1995.
Brief description of amendment: The amendment makes administrative
changes to the KNPP Technical Specifications (TS) to improve their
clarity and consistency. The amendment includes changes to reflect
revisions to 10 CFR Part 20, and changes to correct minor typographical
and format inconsistencies as part of the licensee's ongoing effort to
convert the TS to the WordPerfect format.
Date of issuance: December 21, 1995.
Effective date: December 21, 1995.
Amendment No.: 122.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52936).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1995.
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 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 application for amendments: April 27, 1995, as supplemented
by letter dated November 29, 1995.
Brief description of amendments: The amendments revise TS Table
15.3.5-1, ``Engineered Safety Features Initiation Instrument Setting
Limits,'' and TS Table 15.3.5-3, ``Engineered Safety Features.''
Setting limits are modified and references are changed. The bases for
TS Section 15.3.5, ``Instrumentation System,'' are also changed to be
consistent with the TS changes.
Date of issuance: December 27, 1995.
Effective date: December 27, 1995.
Amendment Nos.: 167 and 171.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27346). The November 29, 1995, submittal provided supplemental
information which did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated December 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Dated at Rockville, Maryland, this 11th day January 1996.
For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV, Office of Nuclear
Reactor Regulation.
[FR Doc. 96-676 Filed 1-19-96; 8:45 am]
BILLING CODE 7590-01-P 11
