[Federal Register Volume 62, Number 98 (Wednesday, May 21, 1997)]
[Notices]
[Pages 27792-27810]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10521]
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 April 28, 1997 through May 9, 1997. The last
biweekly notice was published on May 7, 1997 (62 FR 24984).
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
[[Page 27793]]
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By June 20, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
[[Page 27794]]
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 27, 1996
Description of amendments request: The proposed amendments would
revise Technical Specification (TS) 3.6.1.3.b (peak containment
internal pressure for the design basis loss of coolant accident (LOCA))
from 49.5 psig to 52 psig and the associated Bases Sections. The
proposed amendments reflect values based on a revised LOCA analysis.
The LOCA analysis was revised to reflect the maximum primary
containment internal pressure specified in other TS. This maximum
primary containment internal pressure was not used in the original LOCA
analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff's analysis 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 amendment increases the peak calculated
containment internal pressure for the design basis LOCA from 49.5
psig to 52 psig. The maximum pressure occurs following an accident.
Since the pressure is a consequence of an accident, this change has
no effect on the probability of accident initiation, and therefore,
the probability of an accident previously evaluated has not been
significantly increased.
The consequences of an accident previously evaluated in the
Updated Final Safety Analysis Report (UFSAR) will not be
significantly increased. UFSAR Section 15.6.5.6, ``Analyses of
Effects and Consequences - Large Break LOCA,'' states that ``It is
assumed that the containment leaks at the maximum rates allowed by
the Technical Specifications, i.e., 0.1 vol. %/d for the first 24
hours and half of that rate thereafter.'' The dose calculation
assumes that under accident conditions, the release of radionuclides
to the containment is instantaneously homogenized within the
containment free air volume. This results in a constant
radioactivity per volume (curies/cc) regardless of containment
internal pressure. Since radioactivity is assumed to be homogenized
in the containment free air volume, the volume percent leaked per
day is equivalent to the fraction of radioactivity which leaks from
the containment per day. Therefore, the increase in the peak
calculated containment internal pressure for the design basis LOCA
from 49.5 psig to 52 psig does not effect dose consequences
associated with the design basis LOCA. The proposed change to the
peak calculated containment internal pressure for the design basis
LOCA does not impact the radiological consequences of a LOCA as
analyzed in Chapters 6 and 15 of the UFSAR.
The proposed amendments do not, therefore 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 possibility of a new or different kind of accident has not been
created. The increase in the peak calculated containment internal
pressure for the design basis LOCA does not affect the design or
operation of existing plant equipment, nor involve new plant equipment.
Therefore, the proposed change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The containment design pressure is 60 psig. The acceptance
criteria in NRC Standard Review Plan, Section 6.2.1.1.A, ``PWR Dry
Containments, including Subatmospheric Containments,'' requires in
Item 11.1 that ``the containment design pressure should provide at
least a 10% margin above the accepted peak calculated containment
pressure following a loss of coolant accident.'' For PVNGS to
maintain the required margin, this requires that the peak calculated
containment internal pressure for the design basis LOCA would be no
higher than 54 psig. Since the revised peak calculated containment
internal pressure for the design basis LOCA remains below the 54
psig limit, the proposed change does not involve a significant
reduction in the margin of safety.
Based on this review, it appears that the three standards of 10
CFR50.92(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
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: April 9, 1997
Description of amendment request: The proposed change will extend
the existing Technical Specifications surveillance intervals from 7
days to 14 days for the Channel Functional Tests for the refueling
equipment interlocks and for the one-rod-out interlock. The change will
permit, under most normal circumstances, a complete offloading,
shuffling, or onloading of fuel, without the need to halt refueling
activities solely for the performance of these surveillance tests.
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 extends the Technical Specification
Surveillance Requirement (SR) Frequency for the Channel Functional
Tests (CFTs) for the refueling equipment interlocks and the one-rod-
out interlock. The refueling equipment interlocks and the one-rod-
out interlock are explicitly assumed in the analysis of the control
rod removal error during refueling. Criticality, and therefore,
subsequent prompt reactivity excursions are prevented during the
insertion of fuel, provided all control rods are fully inserted
during the fuel insertion. The refueling equipment interlocks
accomplish this by preventing loading fuel into the core with any
control rod withdrawn, or by preventing withdrawal of a control rod
from the core during fuel loading. The one-rod-out interlock and
adequate shutdown margin prevent criticality by preventing
withdrawal of more than one control rod. With one control rod
withdrawn, the core will remain subcritical, thereby preventing any
prompt critical excursion. The proposed change does not change the
function of any of these interlocks, only the frequency at which the
interlocks undergo channel functional testing. A review of past test
performances has demonstrated that extending the Frequency from 7
days to 14 days will not result in any increase in test failures.
Therefore, the proposed change will not change the ability of these
interlocks to perform when required. Based on this, there can be no
significant increase in the radiological consequences of any
previously evaluated accident since all interlocks will continue to
perform as presently analyzed. Therefore, the proposed change does
not involve a significant increase in the
[[Page 27795]]
probability or consequences of an accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change extends the SR Frequency for performing CFTs
for refueling equipment and one-rod-out interlocks. This change does
not result in a modification to the plant or to the manner in which
the plant is operated. The testing will still demonstrated the
operability of the interlocks. Thus, the interlocks will still
function in the same manner. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed change extends the SR Frequency for performing CFTs
on the refueling equipment and one-rod-out interlocks from 7 days to
14 days. Reviews of past test results indicate that extending the
test interval to 14 days will not result in an increase in the
number of CFT failures for these interlocks. This implies that
extending the SR Frequency to 14 days will not result in an increase
in the amount of time the instrument channels will be inoperable
when required to be operable. Since the proposed change does not
result in any reduction in the amount of time the instrument
channels will be operable, 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: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: April 21, 1997
Description of amendment request: The proposed amendment would
revise the Technical Specifications that would (1) reduce the volume of
borated water in the core flood tank (CFT) from 1040 cubic feet to 940
cubic feet, (2) reduce the surveillance acceptance criteria for the
emergency core cooling system (ECCS) high pressure injection (HPI)
flowrate from 500 gallons per minute (GPM) to 431 GPM, and (3) revise a
limiting condition for operation (LCO) which currently allows either
local or remote manual operability of decay heat valves to delete the
local manual valve operability option.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
1. State the basis for the determination that the proposed
activity will not represent a significant increase in the
probability of occurrence or consequences of an accident.
This TSCR [Technical Specification change request] revises the
LCO for RB [reactor building] sump isolation valves, the LCO for the
core flood tank level, and the surveillance requirement for HPI
injection flow rate. The Core Flood and HPI systems are not actuated
until an event occurs. The CFT level used in the new accident
analysis is that level required to be maintained in the CFT
throughout operation (i.e., pre-accident). The new CFT level does
not prevent safe accident mitigation.
Likewise, the reduced HPI flow cannot cause an event to occur,
and while such flow results in less injection to the RCS [reactor
coolant system] when actuated, this is acceptable as demonstrated in
the LOCA [loss-of-coolant accident] analyses. Changes to the LCO for
the RB sump isolation valves support the safety analysis
assumptions. The action statements related to both the level
requirement and flow rates remain unchanged by this request. The
function, operation and surveillance intervals for the isolation
valves (DH-V-6A/B), the CFT level and HPI injection system are not
changed by this request. Therefore, this activity does not increase
the probability of occurrence of an accident, previously evaluated
in the SAR [safety analysis report].
Reducing the CFT nominal volume and reducing the HPI flow
acceptance criteria in the Technical Specifications will not
increase the radiological consequences of any LOCA evaluated in the
SAR. The results of analyses using the reduced CFT inventory and
reduced HPI flow demonstrate that the consequences are within the
limits of 10 CFR 50.46. No fuel failure in addition to that assumed
in the evaluation of the dose consequences would occur. Therefore,
the radiological consequences would not increase.
The editorial changes described above have no impact upon the
probability of occurrence or consequences of an accident.
2. State the basis for the determination that the activity does
not create the possibility of an accident of a new or different type
than any previously analyzed in the SAR.
This TSCR revises the LCO for RB sump isolation valves, the LCO
for the core flood tank level, and the surveillance requirement for
HPI injection flow rate. This change will not adversely affect the
capability of the emergency core cooling systems in the event of a
LOCA. The function, operation and surveillance intervals for both
the borated water level in the core flood tank, and ECCS systems are
not changed by this request and no physical changes or modifications
are being made to Core Flood and HPI system boundaries. Therefore,
because there are no configuration changes this activity does not
create the possibility of an accident or malfunction of a different
type than previously analyzed in the SAR.
In addition, the editorial changes described above do not create
the possibility of an accident of a new or different type than any
previously analyzed in the SAR.
3. State the basis for the determination that the margin of
safety is not significantly reduced.
This TSCR revises the LCO for RB sump isolation valves, the LCO
for the core flood tank level, and the surveillance requirement for
HPI injection flow rate. No system configuration changes (hardware
modifications) will be made to implement the change request, upon
approval of the license amendment. The action requirements for these
technical specifications have not changed. Actions to be taken if
operability requirements are not met include plant shutdown under
certain conditions.
Furthermore, impact upon the margin to safety is limited because
the results of the LOCA analyses demonstrate that the 10 CFR 50.46
acceptance criteria are met, specifically: the PCT [peak clad
temperature] limit and the core-wide oxidation limit of 1 percent of
the fuel cladding, as identified in the Technical Specification
bases. Hence the margin of safety as defined in the bases of any
technical specification is not significantly reduced or impacted by
the implementation of this change request, or the editorial changes
described above.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Patrick D. Milano, Acting
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: April 22, 1997
Description of amendment request: The proposed amendments would
revise Technical Specifications 5.3.1, Fuel Assemblies, and 6.9.1.6,
Core Operating Limits Report, to allow use of
[[Page 27796]]
an alternate zirconium-based fuel cladding, ZIRLO, and limited
substitution of fuel rods by ZIRLO filler rods.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The methodologies used in the accident analyses remain
unchanged. With the exception of a reduction in the heat flux hot
channel factor (FQ), the operating limits will not be
changed. The proposed changes will not result in any equipment
exceeding its design limits under normal or accident conditions. The
calculated doses presented in the UFSAR will remain bounding. Other
than the changes to the fuel assemblies, there are no physical
changes to the plant associated with this Technical Specification
change. A reload safety analysis will continue to be performed for
each cycle to demonstrate compliance with fuel safety design bases.
VANTAGE+ fuel assemblies with ZIRLO clad fuel rods meet the same
fuel assembly and fuel rod design bases as VANTAGE 5H fuel
assemblies. Since the original design criteria are met, the ZIRLO
clad fuel rods will not be an initiator for any new accident. The
clad material is similar in chemical composition and has similar
physical and mechanical properties to Zircaloy. Thus, cladding
integrity is maintained and the structural integrity of the fuel
assembly is not affected. ZIRLO cladding improves corrosion
performance and dimensional stability. No concerns have been
identified with respect to the mixed core of Zircaloy and ZIRLO clad
assemblies. Also, no concerns have been identified with respect to
the use of an individual assembly containing a combination of
Zircaloy and ZIRLO clad fuel rods.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not result in any equipment exceeding
its design limits under normal or accident conditions. All design
and performance criteria continue to be met and no new failure
mechanisms have been identified. The ZIRLO cladding material offers
improved corrosion resistance and structural integrity.
The proposed changes do not affect the operation of any system
or component in the plant. The safety functions of the related
structures, systems, or components are not changed, nor is the
reliability of any structure, system, or component reduced. The
changes do not affect the manner by which the facility is operated
and do not change any facility design feature, structure, or system.
No new or different type of equipment will be installed. Since there
is no other change to the facility or operating procedures, and the
safety functions and reliability of structures, systems, or
components are not affected, the proposed changes do not create the
possibility of a new accident or an accident different from those
previously evaluated.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
Use of ZIRLO fuel cladding material will not result in any
equipment exceeding its design or licensing bases limits under
normal or accident conditions. VANTAGE 5H reload design and safety
analysis limits are unchanged. For each cycle reload core, the fuel
assemblies will be evaluated using NRC-approved reload design
methods, including consideration of the core physics analysis
peaking factors and core average linear heat rate effects. ZIRLO
fuel assemblies will be assessed for use under conditions consistent
with normal core operating conditions allowed in the Technical
Specifications. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: William D. Beckner
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 request: March 26, 1997
Description of amendment request: The proposed amendment would
modify the technical specifications (TSs) which describe the control
room ventilation system autostart functions.
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:
Per 10 CFR 50.92, the proposed changes do not involve a
significant hazards consideration if the proposed changes do not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated;
2. create the possibility of a new or different kind of accident
from any accident previously evaluated; or
3. involve a significant reduction in a margin of safety.
Criterion 1
These changes are administrative in nature, intended to correct
and clarify the TS description of control room ventilation system
operation. Because no changes to plant operations or physical
changes to the plant will occur due to these changes, they do not
involve a significant increase in the probability or consequences of
a previously evaluated accident.
Criterion 2
Because no changes to plant operations or the physical plant
will occur due to these changes, the changes will not create the
possibility of a new or different kind of accident from any
previously evaluated.
Criterion 3
These changes are administrative in nature, intended to correct
and clarify the present TSs with regard to system operation
descriptions. Thus, the changes involve no reduction in margins 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: 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: 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 request: March 26, 1997
Description of amendment request: The proposed amendment would make
three administrative changes to the technical specifications (TSs)
dealing with a grammatical error, an inadvertently deleted frequency
requirement, and a footnote which is no longer applicable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, the proposed changes do not involve a
significant hazards consideration if the proposed changes do not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated;
[[Page 27797]]
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.
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 TSs do not affect the
assumptions, parameters, or results of any UFSAR accident analysis.
The firstproposed change, ``A'', is a grammatical correction; the
second proposed change, ``B'', reformats the page, and returns a
frequency requirement that, while inadvertently deleted from the
TSs, was still met via procedure; the third proposed change deletes
a footnote which is no longer applicable. As described in Section
II.C. of licensee's application request dated March 26, 1997, a load
drop analysis is not required for single-failure-proof load blocks.
Criterion 2
The proposed changes do not involve physical changes to the
plant or changes in plant operating configuration. The changes
described above are essentially administrative in nature, and thus
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Criterion 3
The proposed changes are essentially administrative in nature.
Per NUREG-0612, single-failure-proof cranes are exempt from the
requirements of a load drop analysis; therefore, there is no
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Gail H. Marcus
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: February 12, 1997
Description of amendment request: The proposed amendment would
change position titles in certain Seabrook Station, Unit No. 1
(Seabrook) Appendix A Technical Specifications (TS) to reflect the
present Seabrook organization, would clarify the approval authority for
the Station Qualified Reviewer Program, and would correct a reference.
Specifically, the proposed amendment would:
1. Change TS 6.0, ``Administrative Controls'' to reflect accurately
the current North Atlantic Management organization, their assigned
duties as previously reported to the NRC, and their proper titles,
2. Corrects an incorrect reference in TS 6.4.3.9.b., and
3. Clarifies the term ``Manager'' in TS 6.4.2, ``Station Qualified
Reviewer Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The changes do not involve a significant increase in the
probabilityor consequences of an accident previously evaluated (10
CFR 50.92(c)(1)) because the proposed changes are merely
administrative or editorial in nature. The proposed changes involve
position title changes to reflect current organization, correct an
incorrect reference, and provide clarification with regard to the
organizational level for certain approvals. The changes do not
affect the manner by which the facility is operated and do not
change any facility design feature or equipment. Since there is no
change to the facility or operating procedures, there is no effect
upon the probability or consequences of any accident previously
analyzed.
B. The changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because they do not affect the manner by which
the facility is operated or involve any changes to equipment or
features which affect the operational characteristics of the
facility. Therefore, no new accident initiator is introduced that
could cause a new or different kind of accident from those
previously evaluated. The proposed changes merely involve position
title changes to reflect current organization, correct an incorrect
reference, and provide clarification with regard to the
organizational level for certain approvals.
C. The changes do not involve a significant reduction in a
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes
do not affect the manner by which the facility is operated or
involve equipment or features which affect the operational
characteristics of the facility.Based on this review, it appears
that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.
NRC Project Director: Patrick D. Milano
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: April 15, 1997
Description of amendment request: The proposed amendment would make
changes to Technical Specification Sections 4.3.3.6 and 4.6.4.1, which
require that the hydrogen monitors be periodically tested.
Specifically, the changes to the surveillances would increase the
testing of the monitor's hydrogen sensor, correct inconsistencies
between surveillances, and make changes to the Bases of the
surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO [Northeast Nuclear Energy Company] has reviewed the
proposed changes in accordance with 10CFR 50.92 and has concluded
that the change does not involve a significant hazards consideration
(SHC). The bases for this conclusion is that the three criteria of
10CFR 50.92(c) are not satisfied. The proposed changes do not
involve [an] SHC because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to Technical Specification Surveillances
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test
once per 92 days on a staggered test basis is consistent with the
design and operation of the hydrogen monitor system. The hydrogen
monitoring system is independent of the reactor coolant system
boundary, has no effect on the probability of occurrence of a loss
of coolant accident and performing surveillance testing does not
significantly increase the probability of an accident previously
evaluated.
The proposed change to Technical Specification Surveillances
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test
will not require the opening of a containment isolation valve and
conducting surveillance testing does not significantly increase the
consequence of an accident previously evaluated.
The proposed change to Technical Specification Surveillances
4.3.3.6 and 4.6.4.1 to change the channel check frequency from once
per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an
analog channel operational test to surveillance 4.3.3.6.2 and make
editorial changes to the surveillances and bases sections are
considered administrative changes. Administrative changes do not
involve a significant increase in the
[[Page 27798]]
probability or consequence of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specification Surveillances
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test do
not add any new equipment to the plant and do not affect the way any
system important to safety is operated either in normal or under
accident conditions.
The proposed changes to Technical Specification Surveillances
4.3.3.6 and 4.6.4.1 to change the channel check frequency from once
per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an
analog channel operational test to surveillance 4.3.3.6.2 and make
editorial changes to the surveillances and bases sections are
considered administrative changes.
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 to Technical Specification Surveillances
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test
will provide assurance of expected instrument performance under
accident conditions and performing surveillance testing do not
involve a significant reduction in a margin of safety.
The proposed changes to Technical Specification Surveillances
4.3.3.6 and 4.6.4.1 to change the channel check frequency from once
per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an
analog channel operational test to surveillance 4.3.3.6.2 and make
editorial changes to the surveillances and bases sections are
considered administrative changes. Administrative changes do not
involve a significant reduction in a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed changes do not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: April 17, 1997
Description of amendment request: The proposed amendment would
modify Technical Specification 3.7.14 by clarifying the actions to be
taken when an area temperature exceeds its temperature limit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO [Northeast Nuclear Energy Company] has reviewed the
proposed change in accordance with 10CFR 50.92 and has concluded
that the change does not involve a significant hazards consideration
(SHC). The bases for this conclusion is that the three criteria of
10CFR 50.92(c) are not satisfied. The proposed change does not
involve [an] SHC because the change would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed change to Technical Specification 3.7.14 will
establish allowable tolerances to ensure that the applicable
systems, structures and components are operated within their
existing design bases.
Technical Specification 3.7.14 specifies the actions to be taken
when an area temperature exceeds its temperature limit. The action
taken is dependent on the amount and duration by which the area
temperature exceeds its limit. Actions are currently specified for
exceeding area temperature by less than 20 deg.F and greater than
20 deg.F for periods less than 8 hours and for periods greater than
8 hours. This change clarifies the actions to be taken when the
temperature exceeds its limit by exactly 20 deg.F or exceed its
limit for exactly 8 hours. It is concluded that this change is a
clarification only in that it causes the more conservative actions
to be taken at greater than or equal to 20 deg.F, or at greater
than or equal to 8 hours.
The proposed change, therefore, does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Establishment of tolerances and clarification of actions at a
specific value does not [ ] change the operation of any system,
structure or component during normal or accident conditions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The change is administrative in nature in that it resolves a
discontinuity in the range of temperatures and in the duration
period above the applicable limit for which action is required.
Establishment of tolerances ensures parameters are set and
maintained within allowable design constraints. Clarification of
applicability for the required actions ensures that action is
proscribed for all possible conditions thereby not permitting
operation outside of allowable design.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed change does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Deputy Director: Phillip F. McKee
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: March 31, 1997
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Sections 3/4.6.5.3.2, ``Filtration,
Recirculation, and Ventilation System (FRVS),'' to (1) provide an
appropriate Limiting Condition for Operation and ACTION Statement that
reflects the design basis for the FRVS, and (2) clarify the manner in
which FRVS testing is performed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or
[[Page 27799]]
consequences of an accident previously evaluated.
The proposed TS revisions involve: 1) no hardware changes; 2) no
significant changes to the operation of any systems or components in
normal or accident operating conditions; and 3) no changes to
existing structures, systems or components. Therefore these changes
will not increase the probability of an accident previously
evaluated. Since the plant systems associated with these proposed
changes will still be capable of: 1) meeting all applicable design
basis requirements; and 2) retaining the capability to mitigate the
consequences of accidents described in the HC [Hope Creek] UFSAR
[Updated Final Safety Analysis Report], the proposed changes were
determined to be justified. As a result, these changes will not
involve a significant increase in 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 changes contained in this submittal will not
adversely impact the operation of any safety related component or
equipment. Since the proposed changes involve: 1) no hardware
changes; 2) no significant changes to the operation of any systems
or components; and 3) no changes to existing structures, systems or
components, there can be no impact on the potential occurrence of
any accident. Furthermore, there is no change in plant testing
proposed in this change request which could initiate an event.
Therefore, these changes will 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 for the TS related to the Filtration
Recirculation and Ventilation System (FRVS) Recirculation Subsystem
provide consistency between the Hope Creek TS and post-accident
descriptions of the FRVS Recirculation Subsystem operation already
contained in the UFSAR and reflected in the Hope Creek SER [Safety
Evaluation Report] (NUREG-1048). PSE&G [Public Service Electric &
Gas] believes that the proposed allowed outage times and ACTION
Statements for the FRVS Recirculation Subsystem: 1) will ensure that
the required minimum number of FRVS recirculation units will be
available to mitigate the consequences of accidents described in the
UFSAR; and 2) provide appropriate direction and time requirements
for placing the unit in a safe shutdown condition when the system is
degraded. Therefore, the changes contained in this request do not
result in a significant reduction in a margin of safety.
The revisions to Surveillance Requirement 4.6.5.3.2.b provide an
accurate and clearly defined basis for performing this surveillance
test. The proposed changes implement PSE&Gs existing interpretation
of the TS requirements and therefore do not alter the manner in
which this surveillance test is currently being performed. PSE&G has
concluded that this surveillance test method appropriately tests the
FRVS Recirculation Subsystem. Since the FRVS recirculation units
will continue to be tested with the heaters: 1) operable; and 2) set
at the demand necessary to ``reduce the buildup of moisture,'' PSE&G
believes that the proposed changes to clarify the TS are justified.
Therefore, the changes contained in this request do not result in 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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: March 31, 1997
Description of amendment request: The proposed amendment would
provide changes to Technical Specification (TS) 2.1.2, ``THERMAL POWER,
High Pressure and High Flow,'' ACTION a.1.c for TS 3.4.1.1,
``Recirculation Loops,'' and the Bases for TS 2.1, ``Safety Limits.''
These changes are being made to implement an appropriately conservative
Safety Limit Minimum Critical Power Ratio (SLMCPR) for all Hope Creek
core and fuel designs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The derivation of the revised SLMCPRs for Hope Creek for
incorporation into the Technical Specifications, and its use to
determine cyclespecific thermal limits, have been performed using
NRC approved methods. Additionally, interim implementing procedures
which incorporate cyclespecific parameters have been used which
result in a more restrictive value for SLMCPR. These calculations do
not change the method of operating the plant and have no effect on
the probability of an accident initiating event or transient.
There are no significant increases in the consequences of an
accident previously evaluated. The basis of the MCPR [Minimum
Critical Power Ratio] Safety Limit is to ensure that no mechanistic
fuel damage is calculated to occur if the limit is not violated. The
new SLMCPRs preserve the existing margin to transition boiling and
the probability of fuel damage is not increased. Therefore, the
proposed change does not involve an 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 contained in this submittal result from an
analysis of the Cycle 7 core reload using the same fuel types as
previous cycles. These changes do not involve any new method for
operating the facility and do not involve any facility
modifications. No new initiating events or transients result from
these changes. Therefore, the proposed Technical Specification
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 margin of safety as defined in the Technical Specification
bases will remain the same. The new SLMCPRs are calculated using NRC
approved methods which are in accordance with the current fuel
design and licensing criteria. Additionally, interim implementing
procedures, which incorporate cyclespecific parameters, have been
used. The MCPR Safety Limit remains high enough to ensure that
greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated, thereby preserving
the fuel cladding integrity. Therefore, the proposed Technical
Specification changes do not involve a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: April 11, 1997
Description of amendment request: The proposed amendments would
change Technical Specification 3.6.2.3, ``Containment Cooling System''
and the
[[Page 27800]]
associated bases. The changes would increase the cooling water flow
rate for the 31-day and 18-month surveillances and specify that during
the 31-day surveillance the fans are started and operated in low speed.
The changes are being proposed to ensure that the cooling water flow
rate and the fan speed being verified are representative of the
Containment Fan Cooling Unit post-accident mode of operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes ensure that the fan speed and cooling water
flow rate being verified is representative of the fan speed and
cooling water flow rate required for the post-accident mode of
operation. The proposed changes affect an accident mitigation system
and are being made to assure that the system is being tested in its
accident mitigation mode. There are no new accident initiators
created by the proposed changes. Therefore, the proposed changes do
not involve a significant increase in the probability of an accident
previously evaluated.
The proposed changes provide assurance that the CFCUs will be
capable of maintaining peak containment pressure and temperature
within design limits by verifying the proper post-accident cooling
water flow to the CFCUs. No physical changes to the plant result
from the proposed changes to the surveillance requirements.
Therefore, the proposed changes do not involve a significant
increase in 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 changes for demonstrating operability of the CFCUs
in the low speed mode, with the required post-accident cooling water
flow rate, are consistent with the existing safety function of the
CFCUs following a Design Basis Accident (DBA). The proposed changes
to the surveillance requirements do not involve any physical changes
to plant components, systems or structures, or the operation of the
CFCUs in the post-accident mode. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes to the surveillance requirements provide
assurance that the CFCUs will perform their intended design function
of maintaining peak containment pressure and temperature consistent
with the current design basis following a DBA by verifying the
proper post-accident cooling water flow to the CFCUs. Since the high
speed and low speed control circuits are independent and there are
separate breakers used to energize the CFCU motors in high and low
speed, the CFCUs would be capable of starting in the low speed mode
following a DBA although the high speed breaker and control circuit
may not be available.
Verification of the post-accident flow rate during the 31 day
surveillance also ensures that the required supporting system,
Service Water, is available for normal operation. To ensure that the
containment air temperature is maintained below the initial
temperature condition assumed in the accident analysis during normal
operation, Technical Specification 3/4.6.1.5 requires verification
of the average containment temperature once every 24 hours in Modes
1 through 4.
The proposed changes to the CFCU surveillance requirements do
not affect the ability of the CFCUs to perform their normal and
post-accident functions. These proposed changes ensure the
verification of the proper post-accident service water flow rate to
the CFCUs. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
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: March 26, 1997
Description of amendment request: The proposed amendment would
revise the Virgil C. Summer Nuclear Station Technical Specifications to
change the definition of ``Core Alteration.'' The proposed definition
will not consider movement of components other than fuel, sources, or
reactivity control components. These proposed changes are technically
consistent with the requirements of NUREG-1431, Revision 1,
``Westinghouse Standard Technical Specifications,'' issued on April 7,
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. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes revise the definition of Core Alteration to be
the movement of fuel, sources, or reactivity control components; and to
delete ``or manipulation'' and ``conservative'' from the text. These
changes do not affect the probability of an accident previously
evaluated. The movement of components other than fuel, sources, and
reactivity control components, within the reactor vessel is enveloped
by the analyzed event. Deleting the words ``or manipulation'' and
``conservative'' from the definition of Core Alteration are
administrative changes and also do not impact initiators of analyzed
events. The only component assumed to be an initiator of an analyzed
event is dropping an irradiated fuel assembly, however, fuel is still
part of the definition. Furthermore, a fuel handling accident is
minimized by administrative controls and physical limitations imposed
on fuel handling operations. The movement of components other than
fuel, sources, and reactivity control components within the reactor
vessel will be controlled under plant administrative controls. This
change has no effect on the boron dilution event because when boron
concentration is below limits, Core Alterations are restricted to
maintain the maximum Shutdown Margin. Movement of other components will
have a negligible impact on core reactivity.
The changes to the definition of Core Alteration do not increase
the consequences of an accident previously evaluated. The accident
analysis assumes an irradiated fuel assembly is dropped with the
consequences well within the 10 CFR 100 limits. The dropping of
other components was not addressed in the plant safety analyses,
however, the analysis of the dropped fuel assembly encompasses other
components. The consequences of a boron dilution event are not
addressed because Core Alterations are not allowed when the boron
concentration is below limits. These changes do not affect the
mitigation capabilities of any component or system nor do they
affect the assumptions relative to the mitigation of accidents or
transients. Therefore, the change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes revise the definition of Core Alteration to be
the movement of fuel, sources, or reactivity control components; and to
delete ``or manipulation'' and ``conservative'' from the text. The
change does not involve a
[[Page 27801]]
significant change in the design or operation of the plant. The changes
do not involve a physical alteration of the plant (no new or different
type of equipment will be installed), or new or unusual operator
actions. The changes will not impose any new or different requirements
or eliminate any existing requirements. The definition of Core
Alteration is being clarified and made consistent with NUREG-1431, Rev.
1. Therefore, the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in margin of
safety?
The proposed changes revise the definition of Core Alteration to
be the movement of fuel, sources, or reactivity control components;
and to delete ``or manipulation'' and ``conservative'' from the
text. The safety analysis assumes an irradiated fuel assembly is
dropped. Controls for handling components other than fuel, sources,
or reactivity control components within the reactor vessel are in
plant administrative controls. The effect of a boron dilution event
on Shutdown Margin is limited due to the requirement to suspend Core
Alterations. The movement of other components have a negligible
impact on core reactivity. No change is being proposed, in the
applicability of the definition, to the movement of components which
factor in the design basis analyses (fuel handling accident).
Deleting the terms ``or manipulation'' and ``conservative'' from the
definition of Core Alteration results in a clarification to the
definition that does not technically alter the meaning. Therefore,
the change does not involve a significant reduction in a margin of
safety
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: F. Mark Reinhart, Acting
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: March 26, 1997
Description of amendment request: The proposed amendment would
revise the Virgil C. Summer Nuclear Station Technical Specifications
(TS), Surveillance Requirement (SR) 4.5.2.a, to add (1) the charging/
high head safety injection (HHSI) pump cross connect valves, and (2)
the charging pump mini-flow header isolation valve, to the SR valve
list. The proposed change is an administrative change to meet the
recommendations of NRC Branch Technical Position (BTP) EICSB 18, which
establishes the acceptability of disconnecting power to electrical
components of fluid systems as one means of designing against a single
failure that might cause an undesirable component action. TS SR 4.5.2.a
includes a list of the required positions of manually-controlled,
electrically-operated valves, and identify those valves to which the
requirements for removal of electrical power is applied in order to
satisfy the single failure criterion.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change adds the charging/HHSI pump cross connect
valves and the charging pump mini-flow header isolation valve to the
ECCS [Emergency Core Cooling System] Subsystems - Tavg
(greater than or equal to) 350 deg.F Technical Specification
Surveillance Requirement. This Surveillance Requirement will require
the valves to be verified open with power to the valve operators
removed once per 12 hours. ... The charging/HHSI pump cross connect
valves and the charging mini-flow header isolation valve are not
initiators of any analyzed event. ... The charging pump/HHSI pump
cross connect valves are being modified to meet the recommendations
of the BTP (including this Technical Specification change). The
charging pump mini-flow header isolation valve meets the
requirements of the BTP except it is not located in the Technical
Specifications. ... Requiring the valves to be verified open with
power removed from the valve operator once per 12 hours does not
affect the assumptions relative to the mitigation of accidents or
transients. This requirement ensures that the valves are in a
position with power removed so that a failure will not occur that
will affect the mitigation of an accident. These valves are required
to be open during a LOCA [loss-of-coolant accident]. This change
will ensure that the valves are open with power removed. Therefore,
the change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does this change create the possibility of a new or different
kind of accident from any accident previously evaluated?
...This change does not involve a significant change in the
design or operation of the plant. This change is a result of BTP
EICSB 18. The charging/HHSI pump cross connect valves are being
modified to have power lockout capability, redundant indication on
the main control board, and be included in the Technical
Specifications. This will ensure that a single failure (hot short in
the controls of either valve) will not cause spurious actuation of
the valves during the injection or recirculation phase of the ECCS.
The charging pump mini-flow header isolation valve meets the
requirements of the BTP except it is not located in the Technical
Specifications. The charging/HHSI pump cross connect valves and
charging pump mini-flow header isolation valve are required to
remain open during a LOCA. This modification will ensure that the
valves will remain open during an accident which requires ECCS
operation. The proposed change will not introduce any new accident
initiators. Therefore, the change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in margin of
safety?
...The ECCS is required to operate upon receipt of a safety
injection signal. The charging/HHSI pump cross connect valves and
the charging pump mini-flow header isolation valve are required to
remain open during ECCS operation. However, a single failure may
cause a spurious actuation (closure) of the valves which could
hinder HHSI flow. The modification to the charging/HHSI cross
connect valves (the addition of a power lockout feature and
redundant position indication) and the added TS Surveillance
Requirement will eliminate this failure scenario and ensure the
valves remain in their safety function position (open). The charging
pump mini-flow header isolation valves already contain a power
lockout feature and redundant position indication. These valves are
being added to the Technical Specifications to meet the requirements
of BTP EICSB 18. Therefore, the change does not involve a
significant reduction in a margin of safety[.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Mark Reinhart, Acting
[[Page 27802]]
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: March 13, 1997 (TS 97-01)
Brief description of amendments: The amendments change the
Technical Specifications by raising the allowable U-235 enrichment, as
specified in Section 5.6.1.2, of fuel stored in the new fuel pit
storage racks from 4.5 to 5.0 weight percent.
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 Sequoyah Nuclear Plant (SQN) in accordance with the
proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the allowed enrichment of new fuel stored
in the new fuel storage racks does not change the criticality
potential with the proposed fuel arrangement requirements for the
storage racks. The potential keff values are maintained
the same as the current TS [Technical Specification] requirements.
In addition, the storage racks are not modified, other than the
locations that cannot be filled with fuel assemblies, and the
processes for loading and unloading fuel in these racks and the
controls for these racks remain the same. Since the keff
limits and operating processes are unchanged by the proposed
revision, there is no increase in the probability of an accident
previously evaluated. Likewise, there is no impact to the
consequences of an accident or increase in offsite dose limits as a
result of the proposed TS change because the criticality
requirements are unchanged and plant equipment will be utilized and
operated without change considering the fuel storage location limits
imposed by this request.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
As stated above, the plant equipment and operating processes
will not be altered by the proposed TS change with the exception of
allowed fuel storage locations in the new fuel storage racks. The
limitations on acceptable fuel storage locations in the racks ensure
that the keff limits are maintained at the same limits as
currently required. TVA has not postulated a criticality event at
SQN for the spent or new fuel storage locations because the design
of the associated storage racks, potential moderation, and TS
allowable fuel enrichments do not support the potential for this
condition. Considering the physical barriers that will be installed
and verified to be in place prior to initial loading of fuel in the
new fuel storage racks, the new fuel storage rack physical
limitations will continue to ensure that criticality events are not
credible for the proposed change. Therefore, this change does not
create the potential for a new accident from any previously
analyzed.
3. Involve a significant reduction in a margin of safety.
The proposed TS change maintains the existing requirements for
criticality by utilizing limited storage locations in the new fuel
pit storage racks. There is no change to operating practices
associated with the use and control of these racks except for the
storage limitations. For these reasons, there will be no reduction
in the margin of safety as a result of implementing the proposed TS
change.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
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: April 4, 1997 (TSCR 197)
Description of amendment request: The proposed amendments revise TS
15.6, ``Administrative Controls,'' and 15.7, ``Radiological Effluent
Technical Specifications,'' to change the corporate officer responsible
for nuclear operations from ``Vice President-Nuclear Power,'' to
``Chief Nuclear Officer,'' and to require that the position be an
officer of the company.
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 Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes are administrative only. There are no
physical changes to the facility or its operation. All Limiting
Conditions of Operation, Limiting Safety System Settings, and Safety
Limits specified in the Technical Specification remain unchanged.
Additionally, there are no changes in the Quality Assurance Program,
Emergency Plan, Security Plan, and Operator Training and
Requalification Program. Therefore, an increase in the probability
or consequences of an accident previously evaluated cannot occur.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes are administrative only. No changes to the
facility structures, systems and components or their operation will
result. The design and design basis of the facility remain
unchanged. The plant safety analyses remain current and accurate. No
new or different failure mechanisms are introduced. Therefore, the
possibility of a new or different kind of accident from any accident
previously evaluated is not introduced.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not involve a significant reduction in
a margin of safety.
The proposed amendments are administrative only. All safety
margins established through the design and facility license
including the Technical Specifications remain unchanged. In
addition, the proposed amendments ensure continued emphasis and
assignment of responsibility for overall nuclear safety. Therefore,
all margins of safety are maintained.
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
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: John N. Hannon
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: April 14, 1997 (TSCR 198)
Description of amendment request: The proposed amendments revise TS
15.3.1, ``Reactor Coolant System,'' to require both reactor coolant
pumps to be operable when the reactor is critical and to require that
the reactor be placed in hot shutdown within 6 hours if one or both
reactor coolant pumps cease operating. This revision eliminates the
current provision which allows single pump operation up to 3.5 percent
power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 27803]]
licensee has provided its analysis of the issue of no significant
hazards consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant increase in
the probability or consequences of an accident previously evaluated.
The amendments proposed eliminate an inconsistency in the
Technical Specifications in a conservative manner. The proposed
changes ensure that required protection functions remain operable in
all required modes of operation. Since the protection functions
remain operable in accordance with existing Technical Specification
requirements and serve to mitigate analyzed events no increase in
the consequences of a previously analyzed accident results. The
protective functions are not accident initiators and are maintained
and tested in accordance with existing Technical Specification
requirements, therefore the probability of a previously analyzed
accident cannot increase. Therefore, operation of the Point Beach
Nuclear Plant in accordance with the proposed changes does not
result in an increase in probability or consequences of a previously
analyzed accident.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendments restore consistency within the Technical
Specifications thus ensuring the protections functions remain
operable as required and the units are operated within the bounds of
the existing safety analyses. Therefore, operation of the Point
Beach Nuclear Plant in accordance with the proposed amendments does
not result in a new or different kind of accident from any accident
previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not involve a significant reduction in
a margin of safety.
Margins of safety are defined by the bounds of the design and in
the safety analyses performed for the Point Beach Nuclear Plant. The
proposed amendments eliminate an inconsistency within the Technical
Specifications and ensure the plant will respond as analyzed in the
Safety Analyses. There is no physical change in the facility or
operation. Therefore, operation of the Point Beach Nuclear Plant in
accordance with the proposed amendments does not involve a reduction
in 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
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: John N. Hannon
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: March 21, 1997, as supplemented by
letter dated April 15, 1997.
Description of amendment request: This amendment request proposes
to revise the technical specifications associated with the inspection
of the reactor coolant flywheel to provide an exception to the
recommendations of Regulatory Guide 1.14, Revision 1, ``Reactor Coolant
Pump Flywheel Integrity.'' The proposed exception would allow either an
ultrasonic volumetric examination or surface examination to be
performed at approximately 10-year intervals. In addition, a correction
of the issuance date of a referenced regulatory guide is included.
This amendment would also allow delaying the complete flywheel
examination for the ``D'' reactor coolant pump until the Fall 1997
outage.
This supersedes the staff's proposed no significant hazards
consideration determination evaluation for the requested changes that
was published on January 2, 1997 (62 FR 133).
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 safety function of the RCP [reactor coolant pump] flywheels
is to provide a coastdown period during which the RCPs would
continue to provide reactor coolant flow to the reactor after loss
of power to the RCPs. The maximum loading on the RCP flywheel
results from overspeed following a LOCA [loss-of-coolant accident].
The maximum obtainable speed in the event of a LOCA was predicted to
be less than 1500 rpm. Therefore, a peak LOCA speed of 1500 rpm is
used in the evaluation of RCP flywheel integrity in WCAP-14535. This
integrity evaluation shows a very high flaw tolerance for the
flywheels. The proposed change does not affect that evaluation.
Reduced coastdown times due to a single failed flywheel is bounded
by the locked rotor analysis, therefore, it would not place the
plant in an unanalyzed condition. Therefore, these 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 amendment does not create the possibility of a new
or different kind of accident from any previously evaluated since
the proposed amendments will not change the physical plant or the
modes of plant operation defined in the facility operating license.
No new failure mode is introduced due to the proposed change, since
the proposed change does not involve the addition or modification of
equipment, nor do they alter the design or operation of affected
plant systems, structures, or components.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are basically unchanged by the
proposed amendment. The results of the flywheel inspections
performed have identified no indications affecting flywheel
integrity. As identified in WCAP-14535, detailed stress analysis as
well as risk analysis have been completed with the results
indicating that there would be no change in the probability of
failure for RCP flywheels if all inspections were eliminated.
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 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
[[Page 27804]]
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.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, located in Grundy County,
Illinois
Date of amendment request: January 24, 1997.
Description of amendment request: The application proposed to
change the Technical Specifications to reflect the installation of new
reactor water level instrumentation for the Emergency Core Cooling
System actuation.
Date of publication of individual notice in Federal Register: April
18, 1997 (62 FR 19143).Expiration date of individual notice: May 19,
1997
Local Public Document Room location: The Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, located in Grundy County,
Illinois
Date of amendment request: March 5, 1997.
Description of amendment request: The application proposed to
remove the Main Steam Line Radiation Monitor High scram and the Main
Steam Line Tunnel Radiation High input to the Main Steam Line Isolation
function requirement from the Technical Specifications (TS). The
proposed changes are a result of a Boiling Water Reactor Owners Group
initiative to minimize inadvertent scrams and Main Steam Isolation
Valve closure due to erroneous radiation monitor actuation.
Date of publication of individual notice in Federal Register: April
18, 1997 (62 FR 19141).Expiration date of individual notice: May 19,
1997
Local Public Document Room location: The Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, located in Rock Island
County, Illinois
Date of amendment request: April 21, 1997
Description of amendment request: The amendments would reflect a
change in the Quad Cities, Unit 2, Minimum Critical Power Ratio (MCPR)
Safety Limit and add the Siemens Power Corporation (SPC) methodology
for application of the Advanced Nuclear Fuel for Boiling Water Reactors
(ANFB) Critical Power Correlation to coresident General Electric fuel
for Quad Cities, Unit 2, Cycle 15, to Technical Specification Section
6.9.A.6.b.
Date of publication of individual notice in Federal Register: April
30, 1997 (62 FR 23499)
Expiration date of individual notice: May 30, 1997
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of application for amendment: March 31, 1997
Brief description of amendment: The proposed amendment would revise
the Ginna Station Improved Technical Specifications to reflect a
planned modification to the spent fuel pool storage racks.Date of
publication of individual notice in Federal Register: April 30, 1997
(62 FR 23502)
Expiration date of individual notice: May 30, 1997
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610
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.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of application for amendments: March 5, 1997, as supplemented
May 9, 1997. The May 9, 1997, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
Brief description of amendments: The amendments incorporate a new
Technical Specification for instrumentation associated with automatic
isolation of a pathway for release of non-condensible gases from the
main condenser.
Date of issuance: May 9, 1997
Effective date: May 9, 1997
Amendment Nos.: 185 and 216
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17224) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 9, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: March 14, 1997
Brief description of amendment: The amendment extends the allowed
outage time for its refueling water storage tank
[[Page 27805]]
while performing surveillance testing of its reactor coolant system
pressure isolation valves (Surveillance 4.4.6.2.2).
Date of issuance: May 6, 1997
Effective date: May 6, 1997
Amendment No. 71
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14459) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 6, 1997.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: April 18, 1997, as supplemented
April 29, 1997.
Brief description of amendment: The amendment approves the
modification to the protection circuitry for emergency diesel
generators. The associated Safety Evaluation delineates the staff's
review and findings that the modification and related Final Safety
Analysis Report (FSAR) changes are acceptable.
Date of issuance: May 8, 1997
Effective date: May 8, 1997
Amendment No. 72
Facility Operating License No. NPF-63. The amendment approves
modification to the protection circuitry for emergency diesel
generators and related FSAR changes.
Date of initial notice and proposed no significant hazards
consideration in Federal Register: (62 FR 19818 dated April 23, 1997).
The notice provided an opportunity to submit comments on the
Commission's proposed no significant hazards consideration
determination. No comments have been received. The notice also provided
for an opportunity to request a hearing by May 23, 1997, but indicated
that if the Commission makes a final no significant hazards
consideration determination any such hearing would take place after
issuance of the amendment.The Commission's related evaluation of the
amendment, finding of exigent circumstances, and final determination of
no significant hazards consideration is contained in a Safety
Evaluation dated
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: November 4, 1996, as
supplemented on December 4, 1996, and March 20, 1997.
Brief description of amendments: The amendments revise the
technical specifications (TS) to permit the removal of containment
tendon sheathing filler grease in up to 35 tendons for Byron, Unit 1,
and Braidwood, Unit 1, in advance of the steam generator replacement
outages. The grease will be removed approximately 6 months prior to the
respective steam generator replacement outages. In addition, in
Amendment No. 80 issued on April 16, 1997, the title in Braidwood's TS
6.9.1.7 was unintentionally left uncorrected. The corrected page is
included in this amendment.
Date of issuance: May 6, 1997
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 89, 89 and 81, 81
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 15, 1997 (62 FR
2186). The March 20, 1997, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 6, 1997No significant hazards
consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
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: February 17, 1997, as
supplemented February 27, March 12, March 26, April 2, and April 10,
1997
Brief description of amendments: The amendments evaluate the
Unreviewed Safety Question (USQ) associated with the use of containment
pressure to compensate for the deficiency in Net Positive Suction Head
(NPSH) for the Emergency Core Cooling System (ECCS) pumps following a
Design Basis Accident (DBA). In the resolution of the USQ, the licensee
changed the Updated Final Safety Analysis Report (UFSAR) in the
following areas:
1. containment analysis,
2. decay heat model,
3. increase in the suppression pool temperature and the effect on
other associated systems following a DBA, and
4. ECCS heat exchanger duty and containment cooling service water
(CCSW) system flow.In addition, the proposed amendments would change
the Technical Specification (TS) allowable water temperature limits for
the suppression chamber and the ultimate heat sink from less than or
equal to 75 degrees Fahrenheit to less than or equal to 95 degrees
Fahrenheit. The original licensing basis water temperature for both the
suppression chamber and ultimate heat sink was 95 degrees Fahrenheit.
Both values were changed in the TS in Amendment Nos. 152 and 147 for
Dresden, Units 2 and 3, respectively, issued on January 28, 1997. The
amendments to lower the ultimate heat sink and suppression pool
temperature limits in the TS was in response to the resolution of a USQ
associated with the operation of Dresden, Units 2 and 3, following the
discovery of a calculational error concerning the head loss across the
ECCS suction strainers. The proposed amendments will return both units
to normal operating conditions allowing for continued power operations
when the ultimate heat sink temperature goes above 75 degrees
Fahrenheit during warm weather.
Date of issuance: April 30, 1997
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 157; 152.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the licenses, TS and USFAR.
Date of initial notice in Federal Register: February 27, 1997 (62
FR 8998). The February 27, March 12, March 26, April 2 and April 10,
1997, submittals provided additional clarifying information that did
not change the initial proposed no significant hazards consideration
determination.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated
[[Page 27806]]
April 30, 1997No 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 Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: February 17, 1997
Brief description of amendments: The amendments would change the
Technical Specifications by increasing the load test values of the
emergency diesel generators in Surveillance Requirement 4.9.A.8.h from
between 2625 kW and 2750 kW to 2730 kW and 2860 kW.
Date of issuance: May 1, 1997
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 176 and 172
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14460). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 1, 1997.No significant
hazards consideration comments received: No
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: August 22, 1996, as supplemented
March 28, 1997.
Brief description of amendment: The amendment revises Technical
Specification Sections 3.3 and 4.5 to allow the deletion of the
requirement to utilize sodium hydroxide (NaOH) as an additive in the
post-accident containment spray system.
Date of issuance: April 23, 1997
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 191
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4345) The March 28, 1997, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the
amendment request as originally noticed. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
April 23, 1997No significant hazards consideration comments received:
No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
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: March 7, 1997, as supplemented
by letters dated April 2, 10, 16, 22, and 28, 1997
Brief description of amendments: The amendment revise Section 3/
4.7.1.6 of the Technical Specifications to require four instead of
three steam generator pressure operated relief valves operable.
Date of issuance: April 29, 1997
Effective date: As of the date of issuance to be implemented within
30 days. Implementation of the amendments include the incorporation in
the Updated Final Safety Analysis Report (UFSAR) of the changes to the
description of the facility as set forth in the licensee's application
dated March 7, 1997, as supplemented by letters dated April 2, 10, 16,
22, and 28, 1997, as evaluated in the staff's Safety Evaluation dated
April 29, 1997.
Amendment Nos.: 159 and 151
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications and License Conditions.
Date of initial notice in Federal Register: March 13, 1997 (62 FR
11931) The April 2, 10, 16, 22, and 28, 1997, letters provided
additional and clarifying information that did not change the scope of
the March 7, 1997, 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 April 29, 1997.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 30, 1994, as
supplemented by letters dated September 18, 1995, and March 15, April
29, May 16, September 23, and October 28, 1996, and January 16, April
22, and May 2, 1997
Brief description of amendments: The amendments revise the
Technical Specifications related to the replacement of the Westinghouse
Model ``D'' type preheat steam generators with feedring steam
generators designed by Babcock and Wilcox International.
Date of issuance: May 5, 1997
Effective date: As of the date of issuance to be implemented within
30 days for Unit 1; and effective upon replacement of the steam
generators for Unit 2.
Amendment Nos.: 175 and 157
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56366) The March 15, April 29, May 16, September 23, and October 28,
1996, and January 16, April 22, and May 2, 1997, letters provided
clarifying information that did not change the scope of the September
30, 1994, 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 May 5, 1997.No significant hazards
consideration comments received: No
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, North Carolina 28223-0001
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 15, 1996
Brief description of amendment: The amemdment revises the technical
specifications to allow the performance of the 24-hour emergency diesel
generator maintenance run while the unit is in either Mode 1 or Mode 2.
Date of issuance: May 5, 1997
Effective date: May 5, 1997
Amendment No.: 94
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
127) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 5, 1997.No significant hazards
consideration comments received. No.
Local Public Document Room location: Government Documents
[[Page 27807]]
Department, Louisiana State University, Baton Rouge, LA 70803
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: April 11, 1997
Brief description of amendment: The amendment would permit steam
generator tubes with intergranular corrosion indications that may
exceed through-wall limits to remain in service until the next
refueling outage.
Date of issuance: May 7, 1997
Effective date: May 7, 1997
Amendment No.: 189
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration (NSHC): Yes (62 FR 19628 dated April
22, 1997). The notice provided an opportunity to submit comments on the
Commission's proposed NSHC determination. No comments have been
received. The notice also provided for an opportunity to request a
hearing by May 22, 1997, but indicated that if the Commission makes a
final NSHC determination, any such hearing would take place after
issuance of the amendment. The Commission's related evaluation of the
amendment, finding of exigent circumstances, and final determination of
NSHC are contained in a Safety Evaluation dated May 7, 1997.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
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: December 19, 1996
Brief description of amendment: The proposed changes revise
Technical Specification Table 4.3-1 to change the power calibration
requirements for the linear power level, the Core Protection Calculator
(CPC) delta T power and the CPC nuclear power signals between 15 and 80
percent power to allow more conservative settings.
Date of issuance May 5, 1997
Effective date: May 5, 1997, to be implemented within 30 days.
Amendment No.: 183
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4348) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 5, 1997.No significant hazards
consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: November 12, 1996, as
supplemented November 27, 1996 (TSCR 224)
Brief description of amendment: The amendment updates the technical
specifications to reflect the implementation of the revised 10 CFR Part
20, ``Standards for Protection Against Radiation.''
Date of issuance : May 8, 1997
Effective date: May 8, 1997, with full implementation within 30
days.
Amendment No.: 191
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66708). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated May 8, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: February 7, 1997
Brief description of amendment: The amendment revises TS 3.12,
``Station Service Power,'' to require both 115 kV power circuits to be
operable when the reactor is critical and to limit or restrict the time
during which Maine Yankee may continue to operate if one or both of the
115 kV power circuits become inoperable.
Date of issuance May 2, 1997
Effective date: May 2, 1997, to be implemented within 30 days.
Amendment No.: 157
Facility Operating License No. DPR-36: Amendment revised the
Technical Specifications and/or License.
Date of initial notice in Federal Register: February 26, 1997 (FR
8799) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 2, 1997No significant hazards
consideration comments received: No
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
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 February 24, 1997, as supplemented by
letters dated March 13, April 11, 23, and 29, 1997
Brief description of amendments: The amendments change the
Technical Specification surveillance requirements for the Control Room
Emergency Filtration System, the Penetration Room Filtration System,
and the Containment Purge Exhaust Filter System.
Date of issuance May 1, 1997
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 127 and 121
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: March 6, 1997 (62 FR
10294) The March 13, April 11, 23, and 29, 1997, letters provided
clarifying information that did not change the scope of the February
24, 1997, 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 May 1, 1997.No
significant hazards consideration comments received: No
Local Public Document Room location: Houston-Love Memorial
Library, 212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama
36302
Tennessee Valley Authority, Docket Nos. 50-260, and 50-296, Browns
Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: June 21, 1996, supplemented
February 7, 1997 (TS 377)
Brief description of amendments: The amendments provide a new
minimum critical power ratio safety limit to replace a nonconservative
value. Technical Specification Bases are also updated to clarify usage
of the residual heat removal system supplemental spent fuel pool
cooling mode.
Date of issuance : May 7, 1997
Effective date: As of the date of issuance to be implemented
within 30 days from the date of issuance.
Amendment Nos.: 247 and 207
[[Page 27808]]
Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register:
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 7, 1997.No significant hazards
consideration comments received: No.
Local Public Document Room location: Athens Public library, 405 E.
South Street, Athens, Alabama 35611
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: September 4, 1996, as
supplemented February 3, 1997. The February 3, 1997 submittal provided
clarifying information only, and did not change the proposed no
significant hazards consideration determination.
Brief description of amendments: The amendments revise the license
and technical specifications (TS) to permit the insertion of four
demonstration fuel assemblies into the reactor core of either North
Anna 1 or North Anna 2, as described in the licensee's submittal. The
four lead test assemblies, fabricated by Framatome Cogema Fuels, will
incorporate several advanced design features, including: a debris
filter bottom nozzle, mid-span mixing grids, a floating top end grid, a
quick disconnect top nozzle, and use of advanced zirconium alloys for
fuel assembly structural tubing and for fuel rod cladding.
Date of issuance May 9, 1997
Effective date: May 9, 1997
Amendment Nos.: 204 and 185
Facility Operating License Nos. NPF-4 and NPF-7. These amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64396) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 9, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: July 18, 1996, as supplemented
on January 29, 1997.
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant Technical Specification 3.8, ``Refueling,'' and its
associated Basis, by allowing the containment personnel air lock doors
to remain open during refueling operations.
Date of issuance May 7, 1997
Effective date: May 7, 1997
Amendment No.: 132
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 14, 1996 (61 FR
42285). The January 29, 1997, submittal provided supplemental
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
May 7, 1997.No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
Notice Of Issuance Of Amendments To Facility Operating LicensesAnd
Final Determination Of No Significant Hazards ConsiderationAnd
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to
[[Page 27809]]
Facility Operating License, 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 room for the
particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By June 20, 1997, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
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 20555-001, 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 20555-001, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Commonwealth Edison Company, Docket No. 50-265, Quad Cities Nuclear
Power Station, Unit 2, Rock Island County, Illinois
Date of application for amendment: April 29, 1997.
Brief description of amendment: The proposed amendment modifies
Section 5.3.A, ``Design Features'' of the Technical Specifications (TS)
to reflect the ATRIUM-9B fuel design and would include various Siemens
Power Corporation (SPC) topical reports in TS Section 6.9.A.6, ``Core
Operating Limits Report,'' to reflect mechanical design criteria for
this fuel and topical reports required for operation. This change would
allow this fuel to be loaded into the core only under Operational Modes
3 (Hot Shutdown), 4 (Cold Shutdown), and 5 (Refueling) and does not
permit startup or power operation using the ATRIUM-9B fuel.
Date of issuance May 2, 1997
Effective date: May 2, 1997
Amendment No.: 173
Facility Operating License No. DPR-30: The amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No.The Commission's related
evaluation of the amendment, finding of emergency circumstances and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated May 2, 1997.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
[[Page 27810]]
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
NRC Project Director: Robert A. Capra
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 2, 1997, as superseded May 5, 1997.
Brief description of amendment: The proposed amendment relocates
and revises the requirements for the control of the setpoint for the
Standby Liquid Control system relief valves. The requirements would be
relocated from Section 4.4.A.2.a and Bases Section 3.4.A of the Cooper
Technical Specifications to the Updated Safety Analysis Report and the
Inservice Testing Augmented Testing Program.
Date of issuance May 9, 1997
Effective date: May 9, 1997
Amendment No.: 176
Facility Operating License No. DPR-46: The amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No.The Commission's related
evaluation of the amendment, finding of emergency circumstances and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated May 9, 1997.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: William D. Beckner
Dated at Rockville, Maryland, this 14th day of May, 1997.
For the Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects III/IV, Office of Reactor
Regulation
[Doc. 97-13190 Filed 5-20-97; 8:45 am]
BILLING CODE 7590-01-F
