[Federal Register Volume 63, Number 126 (Wednesday, July 1, 1998)]
[Notices]
[Pages 35986-36002]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-17352]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 8, 1998, through June 19, 1998. The
last biweekly notice was published on June 17, 1998 (63 FR 33103).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication
[[Page 35987]]
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 July 31, 1998, 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: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power
Plant, Unit 1, Monroe County, Michigan
Date of amendment request: January 28, 1998 (Reference NRC-98-0027)
Description of amendment request: The proposed amendment will
revise Section F and I of the Fermi, Unit 1 Technical Specifications to
include requirements for control of effluents; dose limits; annual
reporting in accordance with requirements of 10 CFR 50.36a; and
numerical guideline criteria based on 10 CFR 50, Appendix I. Also, this
amendment will correct several editorial errors.
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 using the standards in 10 CFR 50.92(c). The licensee's
analysis is presented below:
(1) Does the proposed change significantly increase the
probability or consequences of an accident previously evaluated?
No, the proposed submittal establishes additional requirements
and limits on radioactive effluent releases. No existing
requirements are deleted. For these reasons, this proposed change
will not significantly increase the probability or consequences of
an accident at Fermi 1.
[[Page 35988]]
(2) Will the proposed amendment create the possibility of a new
or different kind of accident from any accident previously analyzed?
No, the addition of requirements for radioactive effluent
releases will not cause a new kind of accident. The additional
requirements involve having a functional waste system with
procedures, submitting an annual report, and restricting the
potential dose to the public from effluents. These changes, in
themselves, do not require a different type of operation of systems.
Any new system installed to enable future discharges will be
evaluated at the time of design.
(3) Will the proposed change significantly reduce the margin of
safety at the facilit y?
No, adding new requirements for radioactive effluents will not
decrease the margin of safety. Since no existing requirements are
being eliminated, this change will not reduce the margin of safety
of the facility.
NRC staff has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 50.92(c) are satisfied.
Therefore, NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esquire, Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Branch Chief: John W. N. Hickey.
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power
Plant, Unit 1, Monroe County, Michigan
Date of amendment request: January 28, 1998 (Reference NRC-98-
0025).
Description of amendment request: The proposed amendment will
revise the Technical Specifications on access controls to provide
flexibility while maintaining similar controls over access. Provisions
will be established for cases where work is performed on the Protected
Area boundary, such that the boundary temporarily will not meet the
Technical Specification criteria. Redundancy between Technical
Specifications will be eliminated. Figure B-1, ``Facility Plan,'' will
be modified to show the buildings within the Protected Area, delete
locations of the Protected Area gates and doors, and delete a building
and equipment outside the Protected Area which are planned to be
removed in the future. Finally, several editorial corrections will be
made.
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 using the standards in 10 CFR 50.92(c). The licensee's
analysis is presented below:
(1) Does the proposed change significantly increase the
probability or consequences of an accident previously evaluated?
The proposed changes do not involve a significant increase in
the probability or consequences of an accident. The proposed changes
all involve access control, the Protected Area boundary, or deletion
of details from a sketch, including a building and equipment planned
for removal, which are outside the Protected Area. The changes still
require control over the gates and doors to the Protected Area and
that only authorized individuals will be issued the Fermi 1 key.
Since the changes do not involve operation of any system,
modifications to any required plant systems, nor eliminate the
requirements for control of the Fermi 1 key and access points, the
probability or consequences of an accident will be unaffected.
(2) Will the proposed amendment create the possibility of a new
or different kind of accident from any accident previously analyzed?
The proposed changes do not create the possibility of a new or
different type of accident from any previously evaluated. The
proposed changes will not lead to any different method of operating
any systems, nor will they create any tests involving plant systems.
The changes only affect the access control requirements, the
Protected Area boundary, and deletion of details from a sketch.
Changes of who issues the key, how doors are secured, provisions for
temporary modifications to the boundary, requirements to observe the
Protected Area boundary if degraded, wording consolidation, and more
accurate building outlines cannot cause a new or different type of
accident. Access points and the Fermi 1 key are still required to be
controlled. The Boilerhouse and main unit output transformer are not
used to support the Fermi 1 nuclear facility. Removal of the
Boilerhouse and main unit output transformer from the drawing will
help facilitate future removal plans, but will not cause a new or
different accident from any previously evaluated, since they provide
no support to the Fermi 1 nuclear facility. For these reasons, the
proposed changes to the access control requirements and Figure B-1
will not create the possibility of a new or different type of
accident.
(3) Will the proposed change significantly reduce the margin of
safety at the facility?
The proposed changes do not involve a significant reduction in
the margin of safety. The changes involve access control, the
Protected Area boundary, and the sketch of the facility. Doors and
gates in the Protected Area boundary will still be required to be
secured when personnel are not inside. The keys will still be
required to be controlled and issued only to authorized personnel.
Compensatory measures will be required if the Protected Area
boundary is degraded such that the requirements are not met.
Therefore, there will not be a significant reduction in the margin
of safety.
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, NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esquire, Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Branch Chief: John W. N. Hickey.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: June 5, 1998 (NRC-98-0067).
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 2.1.2 to incorporate cycle-specific
safety limit minimum critical power ratios (SLMCPRs) for the core that
will be loaded during the upcoming refueling outage and update the
footnote associated with the SLMCPR values to limit applicability of
the SLMCPR values to Cycle 7 operation only.
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 license amendment establishes a revised SLMCPR
value of 1.11 for two recirculation loop operation and 1.13 for
single recirculation loop operation for use during Cycle 7
operation. The derivation of the cycle-specific SLMCPRs was
performed using ``General Electric Standard Application for Reactor
Fuel,'' NEDE-24011-P-A-13; U.S. Supplement, EDE-24011-P--A-13-US,
August 1996; and the ``Proposed Amendment 25 to GE Licensing Topical
Report NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit
MCPR.'' Amendment 25 was submitted by General Electric Nuclear
Energy (GENE) to the NRC on December 13, 1996.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established, consistent with NRC
approved methods, to ensure that fuel performance during normal,
transient, and accident conditions is acceptable.
The probability of an evaluated accident is not increased by
revising the SLMCPR
[[Page 35989]]
values. The change does not require any physical plant modifications
or physically affect any plant components. Therefore, no individual
precursors of an accident are affected.
The proposed license amendment establishes a revised SLMCPR that
ensures that the fuel is protected during normal operation and
during any plant transients or anticipated operational occurrences.
Specifically, the reload analysis demonstrates that a SLMCPR value
of 1.11 (1.13 for single loop operation) ensures that less than 0.1
percent of the fuel rods will experience boiling transition during
any plant operation if the limit is not violated.
Based on (1) the determination of the new SLMCPR values using
conservative methods, and (2) the operability of plant systems
designed to mitigate the consequences of accidents not having been
changed;[,] the consequences of an accident previously evaluated
have not been increased.
Additionally, updating of the footnote on the SLMCPR value in
Technical Specification 2.1.2 to limit the applicability of the
SLMCPR values to only Cycle 7 operation will not increase the
probability or consequences of accidents previously evaluated. The
updating of the footnote on the SLMCPR value in Technical
Specification 2.1.2 is an administrative change that has no effect
on the probability or consequences of accidents previously
evaluated.
Therefore, the proposed TS 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 license amendment involves a revision of the SLMCPR
from 1.09 to 1.11 for two recirculation loop operation and from 1.11
to 1.13 for single loop operation based on the results of analysis
of the Cycle 7 core using the same fuel types as in previous fuel
cycles, and updating of the footnote on the SLMCPR values in TS
2.1.2. Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in the
allowable methods of operating the facility. This proposed license
amendment does not involve any modifications of the plant
configuration or changes in the allowable methods of operation.
Therefore, the proposed TS change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed license amendment establishes a revised SLMCPR
value of 1.11 for two recirculation loop operation and 1.13 for
single recirculation loop operation for use during Cycle 7
operation. The derivation of the cycle-specific SLMCPRs was
performed using ``General Electric Standard Application for Reactor
Fuel,'' NEDE-24011-P-A-13; U.S. Supplement, EDE-24011-P-A-13-US,
August 1996; and the ``Proposed Amendment 25 to GE Licensing Topical
Report NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit
MCPR.'' Amendment 25 was submitted by General Electric Nuclear
Energy (GENE) to the NRC on December 13, 1996. Use of these methods
ensures that the resulting SLMCPR satisfies the fuel design safety
criteria that less than 0.1 percent of the fuel rods experience
boiling transition if the safety limit is not violated. Based on the
assurance that the fuel design safety criteria will be met, the
proposed license amendment does not involve a significant reduction
in a margin of safety.
Additionally, updating of the footnote on the SLMCPR value in TS
2.1.2 will not decrease the margin of safety for accidents
previously evaluated. The updating of the footnote on the SLMCPR
value in Technical Specification 2.1.2 is an administrative change
that does not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: May 27, 1998.
Description of amendment request: The request, if granted, would
modify the Technical Specifications to allow the use of various
controlled shift structures during a 36 to 48 hour work week. The
request will allow the use of up to 12 hour shifts without routine
heavy use of overtime.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments will delete the TS 6.2.2.f. requirement
``. . . to have operating personnel work a normal 8-hour day, 40-
hour week while the plant is operating.'' The proposed change will
allow FPL to implement various controlled shift structures and
durations during a nominal (36 to 48 hours) work week. The proposed
changes will allow the use of up to 12 hour shifts without routine
heavy use of overtime. The TS will continue to require the controls
and guidelines for work hours to be contained in administrative
procedures. The proposed amendments do not involve a change to any
structure, system, or component that affects the probability or
consequences of an accident previously evaluated. The proposed
amendments are administrative in nature and do not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments will not change the physical plant or
modes of plant operation and therefore, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed amendments will not result in the
addition or modification of equipment for any systems, structures,
or components at St. Lucie.
The proposed changes modify the controls on working hours for
operating personnel without significantly changing the hours worked
on a weekly or annual basis, and do not alter the current guidelines
on the use of overtime. The changes are administrative in nature.
Consequently, operation of either unit in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed amendments will delete the TS 6.2.2.f. requirement
``. . . to have operating personnel work a normal 8-hour day, 40-
hour week while the plant is operating.'' The proposed change will
allow FPL to implement various controlled shift structures and
durations during a nominal (36 to 48 hours) work week. The proposed
changes will allow the use of up to 12 hour shifts without routine
heavy use of overtime. The TS will continue to require the controls
and guidelines for work hours to be contained in administrative
procedures. This will result in fewer operating shift-to-shift
turnovers per day and will allow more contiguous days off between
work shifts. The net result of longer work shifts will be more
rested crews with better communications between shifts.
The proposed changes do not alter the current guidelines on the
use of overtime and will not alter the basis for any TS that is
related to the establishment of, or maintenance of, a nuclear safety
margin. Consequently, operation of St. Lucie Units 1 and 2 in
accordance with the proposed amendments will not involve a
significant reduction in a margin of safety.
[[Page 35990]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420
NRC Project Director: Frederick J. Hebdon.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: June 3, 1998
Description of amendment request: The request will modify the
Technical Specifications to provide for the use of an interim periodic
method of monitoring oxygen concentration in the service waste decay
tanks in the event that continuous monitoring capability is lost.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed license amendments are administrative in nature and
will rectify an inconsistency between Surveillance Requirement
4.11.2.5.1 and the UFSAR that was inadvertently created by previous
license amendments. The revisions will reinstate a previously
approved conditional exception to the explicit terms of the
presently stated TS requirement to continuously monitor the waste
gases in the on service Waste Gas Decay Tank, and allow limited
system operation using the laboratory gas partitioner to
periodically analyze gas samples in the event that continuous
monitoring capability becomes inoperable. Limits for potentially
explosive mixtures of waste gases have not been altered, and
explosive gas monitoring instrumentation does not prevent or
mitigate design basis accidents or transients which assume a failure
of or a challenge to a fission product barrier. The proposed
revisions do not involve any change to the plant accident analyses
assumptions, and do not involve accident initiators. Therefore,
operation of either facility in accordance with its proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed license amendments are administrative in nature and
rectify an inconsistency between Technical Specification 4.11.2.5.1
and the UFSAR that was inadvertently created by previous license
amendments. The revisions will not change the physical plant or the
modes of plant operation defined in the Facility Licenses. The
changes do not involve the addition or modification of equipment nor
do they alter the design of plant systems. Therefore, operation of
either facility in accordance with its proposed amendment would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed license amendments are administrative in nature and
rectify an inconsistency between Surveillance Requirement 4.11.2.5.1
and the UFSAR that was inadvertently created by previous license
amendments. The revisions will reinstate a previously approved
conditional exception to the explicit terms of the presently stated
TS requirement to continuously monitor the waste gases in the on
service Waste Gas Decay Tank, and allow limited system operation
using the laboratory gas partitioner to periodically analyze gas
samples in the event that continuous monitoring capability becomes
inoperable. Limits for potentially explosive mixtures of waste gases
have not been altered, and explosive gas monitoring instrumentation
does not prevent or mitigate design basis accidents or transients
which assume a failure of or a challenge to a fission product
barrier. The proposed changes do not alter the basis for any
technical specification that is related to the establishment of, or
the maintenance of, a nuclear safety margin. Therefore, operation of
either facility in accordance with its proposed amendment would not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan.
Date of amendment requests: March 3, 1998.
Description of amendment requests: The proposed amendments would
remove the word ``immediately'' from the Unit 1 hydrogen recombiner
surveillance requirement 4.6.4.2.b.4 and revise the Unit 1 and Unit 2
Technical Specification 3/4.6.4 bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
In accordance with 10 CFR 50.92, the proposed changes do not
involve a significant hazards consideration if the 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
This amendment request does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The change removes an ambiguous word from the technical
specification. It does not physically alter the recombiner, nor does
it adversely impact its operating characteristics.
The resistance to ground test will continue to be used to detect
circuit faults. However, with the removal of the word
``immediately'', it will be possible to conduct the test near the
ambient temperature, the temperature for which the 10,000 ohm
criterion is applicable. The previously observed resistance value
that was lower than 10,000 ohms is not indicative of a faulted
heater circuit. Rather, it is the result of an elevated heater
temperature and the electrical characteristics of the heater's
insulating material, magnesium oxide. Magnesium oxide has a negative
electrical resistance temperature coefficient, and it is not unusual
or unacceptable for the measured insulation resistance to be less
than 10,000 ohms when the heater temperature is elevated.
Criterion 2
This proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The hydrogen recombiner is used to mitigate the consequences of an
accident, and it performs no function during normal operation. The
change to the surveillance requirement removes an ambiguous word and
does not affect the equipment or its installed configuration. No
accident initiators that might be introduced by this change have
been identified.
[[Page 35991]]
Criterion 3
This proposed change does not involve a significant reduction in
a margin of safety. The change removes an ambiguous word from the T/
S. The performance characteristics for the recombiner are not
affected by this change, and no margin of safety is impacted.
The resistance to ground test will continue to be used to detect
circuit faults. However, with the removal of the word
``immediately'', it will be possible to conduct the test near the
ambient temperature, the temperature for which the 10,000 ohm
criterion is applicable. The previously observed resistance values
that were lower than 10,000 ohms are not indicative of a faulted
heater circuit. Rather, they are the result of an elevated heater
temperature and the electrical characteristics of the heater's
insulating material, magnesium oxide. Magnesium oxide has a negative
electrical resistance temperature coefficient, and it is not unusual
or unacceptable for the measured insulation resistance to be less
than 10,000 ohms when the heater temperature is elevated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085
Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive,
Buchanan, MI 49107
NRC Acting Project Director: Dr. Ronald R. Bellamy
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: May 7, 1998
Description of amendment request: The proposed revision to the
Millstone Unit 3 licensing basis would address the addition of the dose
from refueling water storage tank (RWST) back leakage into the design
basis loss-of-coolant accident (LOCA) analysis and Chapter 15 of the
Final Safety Analysis Report (FSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve an SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The RWST is a standby system during normal operation, and
provides the initial makeup water supply for the Emergency Core
Cooling System (ECCS) when actuated in response to a Safety
Injection signal. The RWST supply piping does not interface directly
with the Reactor Coolant System or associated Reactor Coolant
Pressure Boundary piping. All piping, up to and including the last
isolation valve prior to the RWST, is rated for pressure exceeding
RSS [recirculation spray system] pump discharge pressure.
The RWST is a passive tank, vented to atmosphere. Following
swapover to post-LOCA recirculation cooling, the RWST is isolated
and is no longer required for accident mitigation purposes. Back
leakage will collect in the tank and mix with any remaining volume
of water. The temperature of the mixed fluid will not significantly
exceed the ambient temperature of the remaining tank volume due to
the extremely low leakage rates involved. Because the tank is vented
to atmosphere, pressurization of the tank [cannot] occur.
The specific condition of back leakage through the RWST
isolation valves in combination with a motor operated valve failure
does not contribute to the probability of a malfunction previously
evaluated in the Safety Analysis Report. In lines that contain a
motor operated valve and result in back leakage to the RWST, there
exists another valve in series. The other valve is either another
motor operated valve, a check valve, or a manually operated valve.
The most limiting single failure assumed is the failure of the
lowest leakage series valve to close and results in the maximum
calculated leakage rate. Certain ECCS check valves are not subject
to single failure consideration and are therefore credited as the
barrier valve against back leakage.
The back leakage into the RWST results in sump water entering
the RWST when it is at its minimum level. The RWST now becomes a
radioactive source and contributes a shine dose to the surrounding
areas. The increase in dose rates onsite will not prevent operators
from remaining in the control room or from accessing equipment
needed to mitigate the accident.
All piping and valves associated with RWST back leakage are
located in harsh radiation areas. Backflow from RSS could increase
dose rates in the areas where these valves are located. Since these
areas are already classified as harsh radiation environments post
LOCA, additional dose contributions from these pipes would not
adversely impact EEQ [environmental qualification of electrical
equipment] doses to vital equipment located in these rooms. Any
vital equipment located within would continue to perform its safety
function.
The leakage back to the RWST has no effect on the ability of the
RSS pumps to perform their design function. The NPSH [net positive
suction head] required by the RSS pumps is not adversely impacted by
the loss of sump water back to the RWST. The RSS switchover to cold
leg recirculation occurs prior to reaching a minimum level of
392,000 gallons in the RWST. Not counting the reactor coolant system
volume, 774,000 gallons of water is in the sump. QSS [quench spray
system] pumps shut off when the inventory in the RWST decreases to
93,000 gallons. Another 303,000 gallons will reach the sump prior to
QSS shutoff. RWST back leakage displaces approximately 36,000
gallons of sump water back into the RWST at the end of 720 hours,
leaving more than 1,000,000 gallons, not counting RCS [reactor
coolant system] volume, in the sump. When RSS switches over to
recirculation, at least 774,000 gallons of water will remain in the
sump. After 720 hours, more water resides in the sump than when RSS
is started. Therefore minimum NPSH requirements will not be impacted
by this leakage.
Post-LOCA back leakage to the RWST has not previously been
included in the radiological consequence analyses for Millstone Unit
3. Including this source in dose assessment increases the
consequences of the accident. NNECO has tested the associated valves
to establish bounding criteria to be used in the analysis of
potential radiological consequences. The contribution of the RWST
back leakage has been determined to be 2.1 Rem at the LPZ [low
population zone] and 0.9 Rem at the Control Room. When combined with
the present LOCA analysis radiological consequences, the results
remain below the previously analyzed values reported in the FSAR.
All dose estimates reflect the limiting exposure which, in this
case, is Thyroid dose. All resultant doses are less than 10CFR100
and GDC [General Design Criterion] 19 limits to offsite and control
room.
Back leakage to the RWST from the operation of RSS is a result
of a LOCA. It cannot increase the probability of a LOCA. Therefore
RWST back leakage does not increase the probability of an accident
previously evaluated.
Based on the above, the proposed license amendment request 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.
No new condition potentially impacting the ability to mitigate
the accident is created by the back leakage. The low leakage rates
from these valves occurs over [an] extended period of time during
which other makeup water sources can be brought into service to
account for lost inventory, if necessary.
Therefore, the proposed license amendment request 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 resultant dose from back leakage of ECCS valves to the RWST
does not reduce the Margin of Safety. The offsite and control room
doses, with the addition from RWST back leakage, remain below the
licensing
[[Page 35992]]
base dose as listed in the SAR [safety analysis report]. Technical
Specification 6.8.4 defines the basis for the leak reduction
program. The basis for the program is to reduce leakage outside
containment to the maximum extent possible. The Technical
Specifications do not define the maximum amount of leakage or the
origin of the leakage. The addition of the back leakage valves to
the leak reduction program does not reduce the Margin of Safety.
Therefore, the proposed license amendment request does not
involve a significant reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 5, 1998
Description of amendment request: The proposed revision to the
Millstone Unit 3 licensing basis would address a recent steam generator
tube rupture (SGTR) analysis that was determined to be an unreviewed
safety question. The SGTR analyses described in the Final Safety
Analysis Report (FSAR) include an offsite dose analysis and a margin to
overfill analysis. Both of the analyses have been updated. The offsite
dose analysis was updated to reflect a larger capacity for the steam
generator atmospheric dump valve, and the margin to overfill analysis
was updated to reflect a new single failure.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve an SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The FSAR Steam Generator Tube Rupture offsite dose analysis is
being updated to reflect a larger capacity for the steam generator
atmospheric dump valve. The updated analysis, as well as the current
FSAR analysis, postulate the failure, in the open position, of the
steam generator atmospheric dump valve associated with the steam
generator with the ruptured tube. Revising the analyses does not
impact the failure probability of the steam generator atmospheric
dump valve. The SGTR analyses credit closure of the atmospheric dump
valve block valve to isolate the failed open atmospheric dump valve.
The revised SGTR analysis uses a larger flow capacity for the
atmospheric dump valve. A larger flow capacity, without other
changes being made, would increase the consequences associated with
this failure. However, the time credited for closure of the block
valve is being reduced to 20 minutes after the atmospheric dump
valve fails open, instead of 30 minutes after the atmospheric dump
valve fails open. A shorter isolation time, without other changes
being made, would decrease the consequences associated with the
atmospheric dump [valve] failing open. This faster isolation time
more than compensates for the larger capacity assumed for the
atmospheric dump valve. Therefore, the revised analyses does not
increase the consequences of a Steam Generator Tube Rupture. The
change is a revision to the analyses for a steam generator tube
rupture and the description of the analyses in the FSAR. Changing
the analyses and its description [cannot] cause an increase in the
probability of a steam generator tube rupture.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The change is to the analyses and FSAR description of that
analyses. The important changes in the analyses are the increased
capacity of the atmospheric dump valve and the shorter time utilized
for isolation of the failed open atmospheric dump valve. The only
change in equipment credited in the analyses is the crediting of the
block valve to close when there is a larger flow through the valve.
The block valve can close under the postulated accident conditions.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The revised analyses reduces the time available to the Operators
to isolate the failed open atmospheric dump valve from 30 minutes to
20 minutes. The actions required are unchanged. The twenty minutes
allows sufficient time for the Operators to both recognize the
failure of the atmospheric dump valve and to close the block valve.
However, reducing the available time to the Operators from 30
minutes to 20 minutes represents a reduction in the margin for error
available to the Operators and thus represents a reduction in the
margin of safety. The reduction in the margin of safety is not
significant since the twenty minutes allowed by the analysis is
still significantly above the typical ten minute minimum assumed
response time for Operator actions performed in the control room. In
addition, Operator training provides assurance that the twenty
minute time limit is met.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 6, 1998.
Description of amendment request: The proposed revision to the
Millstone Unit 3 licensing basis relates to operation of the
supplementary leak collection and release system (SLCRS) after a
postulated accident. Specifically, the proposed revision to the Final
Safety Analysis Report (FSAR) would address (1) the manual actions
required to trip the non-nuclear safety grade fans and time
requirements for control room ventilation realignment, and (2) the
input assumptions and results of the new loss-of-cooling accident/
control rod ejection accident analyses.
Basis for proposed no significant hazards consideration
determination:
[[Page 35993]]
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The potential condition of radioactive effluent bypassing the
isolated boundary in the Supplemental Leak Collection and Release
System after an accident cannot contribute to the probability of an
accident previously evaluated. The leakage is caused by a postulated
failure of the non-nuclear safety grade exhaust fans within the
SLCRS boundary to trip after a safety injection signal. Operator
action is needed to verify that the fans in question are tripped
within a predetermined time delay after the accident in order that
credit can be taken in the radiological dose analysis for the
isolation of this source.
The proposed operator action will verify that the power to the
fan motors is terminated, which cannot create any conditions leading
to a new accident. The verification will augment the procedure to
minimize the consequences of the accident itself. The trip circuits
of the fan motors do not interface with safety systems.
The consequences of the limiting design basis accidents have
been evaluated with the additional bypass leakage. The doses for the
Exclusion Area Boundary, Low Population Zone and Unit 3 Control room
remain below the previously calculated and approved licensing
values. The calculated doses for the Technical Support Center are
higher than previously approved, but below the radiological
acceptance criteria of GDC [General Design Criterion] 19.
Therefore, the proposed license amendment 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.
There are no conceivable conditions, created by the proposed
operator action, that may lead to the possibility of a new accident.
Interruption of power to the exhaust fans is, in itself, a part of
accident mitigating activity. The proposed activity cannot create an
adverse environment where a possibility of a new accident has to be
considered.
The breakers used to de-energize the fans, control only the fan
motors and no other equipment. Clear labeling ensures that no safety
equipment is inadvertently de-activated. The revised ventilation
system operating procedure will clearly specify the order of steps
and confirmatory indicators necessary for safe shutdown of the
exhaust fans. The equipment operator will be briefed before
proceeding to open the breakers to the affected fan motors. To
minimize the possibility of an error, this step will be done early
in the sequence of procedural steps performed to re-align the
control room ventilation system to the filtration/recirculation mode
of operation after an accident.
Therefore, the proposed license amendment 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.
In considering the impact of the proposed revision on the margin
of safety, as defined in the Technical Specifications, the impact on
the design basis analysis of the fission product barriers must be
evaluated.
The proposed operator action to trip the fans is done as part of
personnel protective actions after a major accident, which is to
stop the distribution of radioactive iodine into the vital areas
through the ventilation system within a predetermined time. The
maintenance of the fission product barriers is not affected by this
action. This potential source of radioactivity associated with the
ventilation fans discharging through the closed SLCRS boundary
dampers has not been considered previously in the dose analysis.
Including this source results in a small increase in the gamma and
beta doses to the Technical Support Center. The GDC 19 limits for
protection of personnel in the vital areas however, are not
violated. The calculated doses to EAB/LPZ [exclusion area boundary
and the low population zone] zones and to the control room vital
area remain below the current licensing base values.
Therefore, the proposed license amendment request does not
involve a significant reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388;
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: April 23, 1998.
Description of amendment request: The amendment would update the
operating licenses such that the corporate name of Pennsylvania Power
and Light Company ``be changed to PP&L, Inc.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. This request involves an administrative change only. The
Operating Licenses (OLs) are being changed to reference the new
corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed changes.
Therefore, this request will have no impact on the possibility of
any type of accident: new, different, or previously evaluated.
2. Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. This request involves an administrative change only. The OLs
are being changed to reference the new corporate name of the
licensee. No actual plant equipment or accident analyses will be
affected by the proposed change and no failure modes not bounded by
previously evaluated accidents will be created. Therefore, this
request will have no impact on the possibility of any type of
accident: new, different, or previously evaluated.
3. Will the change involve a significant reduction in a margin
of safety?
No. Margin of safety is associated with confidence in the
ability of the fission product barriers (i.e., fuel and fuel
cladding, Reactor Coolant System pressure boundary, and containment
structure) to limit the level of radiation dose to the public. This
request involves an administrative change only. The OLs are being
changed to reference the new corporate name of the licensee.
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed change will not
relax any criteria used to establish safety limits, will not relax
any safety systems settings, or will not relax the bases for any
limiting conditions of operation. Therefore, this request will not
impact 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
[[Page 35994]]
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.
Philadelphia Electric Company, Docket No. 50-171, Peach Bottom Atomic
Power Station, Unit 1, York County, Pennsylvania
Date of application for amendment: March 2, 1998
Brief description of amendment: This proposed amendment will revise
the Peach Bottom Atomic Power Station, Unit 1, Technical Specifications
(TS) to include requirements for control of effluents and annual
reporting in accordance with the requirements of 10 CFR 50.36a.
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. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed changes do not impact the SAFSTOR status of
Unit 1 or the design of any plant system, structure, or component
(SSC). These changes are administrative in nature. They do not
affect security at Unit 1 or the potential of radioactive material
being released. Inspections for potential liquid and gas effluents
have previously been established. These changes ensure the
requirement for procedures and reporting are listed in TS.
Therefore, these proposed changes do not increase the probability or
consequences of an accident previously evaluated.
b. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because implementation of the proposed changes do not involve any
physical changes to plant SSC or impact the SAFSTOR status. The
changes are administrative in nature. Therefore, the possibility of
a new or different kind of accident from any accident previously
evaluated is not created.
c. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed changes do not involve a significant reduction in a
margin of safety because the proposed changes do not affect the
plant SAFSTOR status. Because proposed changes are administrative in
nature, they do not involve a question of safety. These changes
involve reporting and adding a requirement that procedures be in
place for effluent monitoring. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101.
NRC Branch Chief: John W. N. Hickey.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: May 13, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.10.8, ``Inservice Leak and
Hydrostatic Testing,'' to delete the requirement for an operable High
Drywell Pressure trip function. Specifically, TS 3.10.8.a is being
revised to remove the reference to the Secondary Containment Isolation
Actuation Instrumentation trip function 2.b.
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 proposed TS revisions will continue to allow the performance
of inservice leak and hydrostatic testing at a reactor coolant
temperature of greater than 200 degrees Fahrenheit but less than or
equal to 212 degrees Fahrenheit while considering the plant to
remain in Operational Condition 4; however, the requirement to have
an operable ``High Drywell Pressure'' Secondary Containment
Isolation trip function during a leak or hydrostatic test is being
deleted. This change will not have an impact on the consequences of
an accident previously evaluated since the tests will continue to be
performed nearly water solid and with all control rods fully
inserted. The stored energy in the reactor core and coolant will
continue to be very low and the potential for causing fuel failures
with a subsequent increase in coolant activity will continue to be
minimal. The remaining restrictions provided in Special Test
Exception 3.10.8 requiring Secondary Containment Integrity and
Filtration, Recirculation and Ventilation System (FRVS) operability
will continue to provide assurance that potential releases into
secondary containment will be restricted from direct release to the
environment. With the reactor coolant continued to be limited to 212
degrees Fahrenheit, there will be little or no flashing of coolant
to steam, and any release of radioactive materials will be
minimized.
In the event of a large primary system leak, the reactor vessel
will rapidly depressurize, allowing the low pressure Emergency Core
Cooling Systems (ECCS) to operate. The capability of the required
ECCS in Operational Condition 4 remains adequate to maintain the
core flooded under these conditions. Small system leaks will
continue to be detected by leakage inspections, which are an
integral part of the inservice leak and hydrostatic testing
programs, before any significant inventory loss can occur. In
addition, the ``High Drywell Pressure'' Secondary Containment
Isolation trip function (TS Table 3.3.2-1, Trip Function 2.b)
provides no additional protection against the events of concern
during the inservice leak and hydrostatic tests. As a result, these
changes will not increase the probability of an accident previously
evaluated nor significantly increase the consequences of an accident
previously evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to Special Test Exception 3.10.8 contained
in this submittal will not adversely impact the operation of any
safety related component or equipment. Since the proposed changes
involve no hardware changes and no changes to existing structures,
systems or components, there can be no impact on the potential
occurrence of any accident due to new equipment failure modes. The
remaining restrictions provided in proposed Special Test Exception
3.10.8 requiring Secondary Containment Integrity and Filtration,
Recirculation and Ventilation System (FRVS) operability will
continue to function as required, which will provide assurance that
potential releases into secondary containment will be restricted
from direct release to the environment. Furthermore, there is no
change in plant testing proposed in this change request that 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 TS revisions will still allow the performance of
inservice leak and
[[Page 35995]]
hydrostatic testing at a reactor coolant temperature of greater than
200 degrees Fahrenheit but less than or equal to 212 degrees
Fahrenheit while considering the plant to remain in Operational
Condition 4; however, the requirement to have an operable ``High
Drywell Pressure'' Secondary Containment Isolation trip function
during a leak or hydrostatic test is being deleted. Since the
reactor vessel head will remain in place, secondary containment will
continue to be maintained, sufficient isolation actuation
instrumentation will be maintained and all systems required in
Operational Condition 4 will continue to be operable in accordance
with the TS, the proposed changes will not have any significant
impact on any design basis accident or safety limit. Since Hope
Creek will still remain capable of meeting all applicable design
basis requirements and retaining the capability to mitigate the
consequences of accidents described in the UFSAR, the proposed
changes contained in this submittal were determined to 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, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Robert A. Capra.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: June 12, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Limiting Condition for Operation
(LCO) sections 3.7.1.1, 3.7.1.2, and 3.7.1.3. Specifically, the
proposed changes implement more appropriate Ultimate Heat Sink (UHS)
limits for river water temperature, which increases operational
flexibility. In addition, the Station Service Water System (SSWS) and
Safety Auxiliaries Cooling System (SACS) TS Action Statements are being
revised to provide additional restrictions on continued plant
operation. These revisions provide explicit TS guidance, which
maintains SSWS/SACS operating configurations within design analysis
assumptions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
LCO 3.7.1.3 Changes
The proposed TS revisions related to UHS involve no hardware
changes and no changes to existing structures, systems or
components. The UHS and supported system temperature and
configuration limits ensure that the UHS can remove required heat
loads during design basis accidents and transients with the proposed
UHS river water temperature limits. The proposed UHS TS ACTION
Statements ensure that the plant is directed to enter a safe
shutdown condition whenever the capability to mitigate design basis
accidents and transients is lost. The existing UHS TS surveillance
requirements to increase monitoring of the river water temperature
at 82 deg.F adequately ensures that the actions required at elevated
river water temperature conditions are taken as appropriate. Since
the UHS will still remain capable of meeting all applicable design
basis requirements and retaining the capability to mitigate the
consequences of accidents described in the [Hope Creek] HC [Updated
Final Safety Analysis Report] UFSAR, the proposed changes were
determined to be justified. As a result, these changes will not
increase the probability of an accident previously evaluated nor
significantly increase the consequences of an accident previously
evaluated.
LCO 3.7.1.1 and 3.7.1.2 Changes
The proposed TS revisions related to SSWS/SACS operating
configuration restrictions involve no hardware changes and no
changes to existing structures, systems or components. The
additional restrictions requiring: 1) SACS heat exchanger
operability in one SSWS/SACS pump per loop scenarios; and 2)
assessments of SACS loop operability when a SSWS loop is declared
inoperable; ensure that the SSWS/SACS can remove required heat loads
during design basis accidents and transients with the proposed UHS
river water temperature limits contained in this submittal. The
proposed SSWS/SACS TS ACTION Statements ensure that the plant is
directed to enter a safe shutdown condition whenever the capability
to mitigate design basis accidents and transients is lost. Since
SSWS/SACS will still remain capable of meeting all applicable design
basis requirements and retaining the capability to mitigate the
consequences of accidents described in the HC UFSAR, the proposed
changes were determined to be justified. As a result, these changes
will not increase the probability of an accident previously
evaluated nor significantly increase the consequences of an accident
previously evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
LCO 3.7.1.3 Changes
The proposed changes to the UHS TS contained in this submittal
will not adversely impact the operation of any safety related
component or equipment. Since the proposed changes involve no
hardware changes and no changes to existing structures, systems or
components, there can be no impact on the potential occurrence of
any accident due to new equipment failure modes. The system
configuration limits imposed by the UHS LCO ensure that supported
systems can remove required heat loads during design basis accidents
and transients with the proposed UHS river water temperature limits.
Furthermore, there is no change in plant testing proposed in this
change request that 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.
LCO 3.7.1.1 and 3.7.1.2 Changes
The proposed changes to the SSWS/SACS TS contained in this
submittal will not adversely impact the operation of any safety
related component or equipment. Since the proposed changes involve
no hardware changes and no changes to existing structures, systems
or components, there can be no impact on the potential occurrence of
any accident due to new equipment failure modes. The system
configuration limits imposed by the SSWS/SACS LCOs ensure that
systems can remove required heat loads during design basis accidents
and transients with the proposed UHS river water temperature limits.
Furthermore, there is no change in plant testing proposed in this
change request that 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.
LCO 3.7.1.3 Changes
The proposed changes for the TS related to the UHS ensure
continued capability of the UHS to mitigate the consequences of
design basis accidents and transients. The UHS supported systems'
configuration limits and changes to the operating limits of the UHS
ensure that the UHS can remove required heat loads during design
basis accidents and transients with the proposed river water
temperature limits. The proposed UHS TS ACTION Statements ensure
that the plant is directed to: 1) enter a safe shutdown condition
whenever the capability to mitigate design basis accidents and
transients is lost; or 2) enter a conservatively short period of
continued operation when supported system redundancy is reduced.
Since the UHS will still remain capable of meeting all applicable
design basis requirements and retaining the capability to mitigate
the consequences of accidents described in the HC UFSAR, the
proposed changes contained were determined to not result in a
significant reduction in a margin of safety.
LCO 3.7.1.1 and 3.7.1.2 Changes
The proposed changes for the TS related to the SSWS/SACS ensure
continued capability of these systems to mitigate the consequences
[[Page 35996]]
of design basis accidents and transients. The proposed configuration
limits ensure that the safety-related heat removal systems can
perform their safety functions during design basis accidents and
transients with the proposed river water temperature limits. The
SSWS/SACS TS ACTION Statements ensure that the plant is directed to:
1) enter a safe shutdown condition whenever the capability to
mitigate design basis accidents and transients is lost; or 2) enter
a conservatively short period of continued operation when supported
system redundancy is reduced. Since the SSWS/SACS will still remain
capable of meeting all applicable design basis requirements and
retaining the capability to mitigate the consequences of accidents
described in the HC UFSAR, the proposed changes contained were
determined to 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, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Robert A. Capra.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 7, 1998.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to reflect reactor coolant
system flow differences between the existing Model E and the
replacement Delta 94 steam generators (SGs). Specifically, it would (1)
add a new reactor core safety limit figure in TS 2.1.1, Reactor Core
Safety Limits, that shows curves that are a function of core
temperature, power and operating pressure, applicable to the Delta 94
SGs, (2) add a footnote in TS Table 2.2-1, Reactor Trip System
Instrumentation Trip Setpoints, to specify a new design loop flow rate
applicable to the Delta 94 SGs, and (3) add a new flow rate requirement
to TS 3.2.5, Departure from Nucleate Boiling (DNB) Parameters,
applicable to the Delta 94 SGs. Related changes to the TS Bases were
also proposed for Bases 2.1.1, Reactor Core Safety Limits, and Bases 3/
4.2.5, DNB Parameters.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed Technical Specification changes are necessary to
reflect new conditions associated with replacement of the steam
generators. The differences in the replacement steam generators only
require small changes to parameters modeled in existing accident
analyses. Accident analyses affected by the replacement steam
generator parameter changes have each been evaluated to establish
that there is no significant change in the documented results. In
cases where an evaluation was not adequate, new analyses have been
performed to verify that there is no significant change in the
consequences of the affected accidents.
The Technical Specification changes specify new requirements
(i.e., changed RCS [reactor coolant system] flow) which support the
new and existing accident analyses. The accident analysis performed
for these new requirements determined that neither the probability,
nor the consequences, of accidents previously evaluated in the UFSAR
[Updated Final Safety Analysis Report] would be increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed Technical Specification changes are necessary to
reflect new conditions associated with replacement of the steam
generators. The differences in the replacement steam generators only
require small changes to parameters modeled in existing accident
analyses. The replacement of the original steam generators with new
Model Delta 94 steam generators improves the structural integrity of
the steam generator tubes. The improved structural integrity of the
new steam generators does not increase the possibility of a new or
different kind of accident from any accident previously evaluated
such as a multiple steam generator tube rupture event.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not alter the manner in which Safety
Limits, Limiting Safety System Setpoints, or Limiting Conditions for
Operations are determined. Changes in parameters assumed in safety
analyses associated with replacement of the steam generators have
been analyzed and new Technical Specification limits are proposed.
The new limits proposed for SL [Safety Limit] 2.1.1, ``Reactor
Core''; Table 2.2-1, ``Reactor Trip System Instrumentation Trip
Setpoints''; and LCO [Limiting Condition for Operation] 3.2.5, ``DNB
[Departure from Nucleate Boiling] Parameters'' maintain or improve
the 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: John N. Hannon.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: June 1, 1998.
Description of amendment request: The proposed amendment would
revise the minimum steam generator (SG) tube roll expansion distances
for the F* and elevated F* (EF*) repair criteria that were approved in
Amendment 129.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
(1) Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The changes to the minimum engagement lengths for F* and EF* do
not change any of the conclusions of the original F* and EF*
analyses. The technical justification for the repair criteria has
not changed due to changes in the engagement lengths. The calculated
engagement lengths continue to preclude tube pullout and rupture
during all postulated conditions. Based on the geometry of the Model
51 SG, tube rupture type release rates are not expected for a
postulated failure at an F* or EF* repair location. Engagement
lengths were calculated such that structural integrity of the
repaired tube meets the RG [Regulatory Guide] 1.121 requirements.
Therefore, application of the new F* and EF* distances will not
increase the probability of an accident previously evaluated.
The new calculated engagement lengths continue to preclude
primary to secondary leakage during all conditions. Leakage for both
F* and EF* remains negligible at normal operating conditions. The
amount of leakage expected at faulted conditions from F* and EF*
repaired tubes remains a small percentage of the maximum allowable
leak rate during a[n] SLB [steamline break] and is considered
negligible. Therefore, it can be concluded that leakage will be
restricted such that off-site doses will not exceed a small fraction
of 10 CFR part 100 and control
[[Page 35997]]
room doses will not exceed GDC [General Design Criterion] 19
criteria. Therefore, the proposed change to the F* and EF* distances
will not increase the consequences of an accident previously
evaluated.
(2) Create the possibility of a new or different kind of
accident from any previously evaluated.
Implementation of the proposed changes in F* and EF* distances
does not introduce any significant changes to the plant design
basis. As with the original acceptance of the amendment for using
the original F* and EF* criteria, use of the proposed F* and EF*
engagement lengths will not introduce a mechanism that will result
in an accident initiated outside of the tubesheet crevice region. As
previously discussed, the structural integrity of F* and EF* tubes
will be maintained during all plant conditions. Any hypothetical
accident as a result of tube degradation in the tubesheet crevice
region of the tube will be bounded by the existing tube rupture
analysis. Therefore, implementation of the proposed engagement
lengths for F* and EF* will not create the possibility of a new or
different kind of accident.
(3) Involve a significant reduction in the margin of safety.
The calculation for the new F* and EF* minimum engagement
lengths used the same methodology as the original F* and EF*
analysis. The only change was the assumed normal operating primary
to secondary differential pressure. The new assumed differential
pressure is the design differential pressure for the KNPP [Kewaunee
Nuclear Power Plant] SGs. The calculation for the engagement lengths
continues to use the appropriate safety factors from RG 1.121. The
revised F* and EF* engagement lengths continue to preclude tube
pullout at all plant conditions and to maintain the structural
integrity of the tube. Additionally, primary to secondary leakage
during all plant conditions is precluded as described in the
preceding sections. Since the structural and leakage integrity is
not changed by the proposed changes in engagement length, the margin
of safety is not significantly reduced.
Additionally, use of the F* and EF* repair criteria will
decrease the number of tubes removed from service by plugging or
repaired by sleeving. Since both plugging and sleeving reduce
reactor coolant flow margin, implementation of the F* and EF* repair
criteria helps to maintain that flow margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Acting Project Director: Ronald R. Bellamy.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: February 2, 1998, as supplemented
February 18, 1998.
Description of amendment request: The proposed amendments would
revise the wording to specify refueling outage surveillances. The
changes clarify that these surveillances are to be performed on an 18-
month frequency and need not be constrained to refueling outage
conditions.
Date of publication of individual notice in Federal Register:
February 10, 1998 (63 FR 6784).
Expiration date of individual notice: For comments February 24,
1998; For hearing March 12, 1998.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Duke Energy Corporation, Docket Nos. 50-269 and 50-287, Oconee Nuclear
Station, Units 1 and 3, Oconee County, South Carolina
Date of amendment request: June 4, 1998.
Description of amendment request: The proposed amendments would
revise Technical Specification 4.17.2 to allow continued operation with
certain steam generator tubes that exceed their repair limit as a
result of tube end anomalies. This action temporarily exempts these
tubes from the requirement for sleeving, rerolling, or removal from
service until they are repaired during or before the next scheduled
refueling outages for the respective unit.
Date of publication of individual notice in Federal Register: June
17, 1998 (63 FR 33097).
Expiration date of individual notice: For comments July 1, 1998;
For hearing July 17, 1998.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of application for amendments: June 6 and December 11, 1996,
April 11, May 1, August 14, October 15, November 5 and 14, December 3,
4, 15, 22, 23, 29 and 30, 1997, January 23, March 12 and 13, April 16,
20 and 28, May 7, 14 and 19, and June 2, 1998.
Brief description of amendments: Conversion to Standard Improved
Technical Specifications (TSs). Supplements requested less restrictive
changes to the planned conversion. These changes involve (1) plant-
specific application of generically approved methodology supporting
extended instrument surveillance intervals and allowed outage times,
(2) operating practice to treat secondary containment as a single zone,
(3) TS changes to support installation of a Power Range Neutron
Monitoring System, Average Power Range Monitor and Rod Block Monitor TS
improvements, and the Maximum Extended Load Line Limit analysis, (4)
TSs to specify reactor vessel water level should be greater than the
top of the irradiated fuel, (5) reflect plant-specific design condition
that excludes average U-235 enrichment, (6) all spiral off-load
procedures and adopt revision to Surveillance Requirement (SR). Also,
changes to (1) SR relating to core reactivity difference between actual
and expected critical rod configuration, (2) calibration frequency for
local power range monitors and (3) an alternate SR for Unit 3 for
position verification of the low pressure core injection cross tie
valves.
Date of publication of individual notices in the Federal Register:
June 1, 1998 (63 FR 29763), and June 12, 1998 (63 FR 32252).
Expiration dates of individual notices: July 1, 1998 (63 FR 29763)
and July 13, 1998 (63 FR 32252).
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
[[Page 35998]]
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: October 1, 1997, as
supplemented October 14, 1997, March 16, April 1 and 28, May 1 and 20,
1998.
Brief description of amendments: Change Technical Specifications to
allow operation at the uprated power level of 3458 MWt which represents
a power level increase of 5 percent.
Date of publication of individual notice in the Federal Register:
June 9, 1998 (63 FR 31533).
Expiration date of individual notice: July 9, 1998.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
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.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: March 25, 1998, as supplemented
on April 8, and May 5, 1998.
Brief description of amendment: The amendment modifies the Pilgrim
Nuclear Power Station Technical Specification Section 3.6.A.1 with
respect to the monitoring requirements for the vessel flange and
adjacent shell differential temperature during heatup and cooldown and
removes the 145 deg.Fahrenheit differential temperature limit.
Date of issuance: June 19, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 175.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1998 (63 FR
23304). The May 5, 1998, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 19, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: July 15, 1996, as supplemented
on June 19, 1997, and February 2, 1998.
Brief description of amendments: The amendments relocate
requirements related to fire protection from the Technical
Specifications (TS) to the Updated Final Safety Analysis Report. The TS
sections to be relocated are: 3/4.3.7.9, Fire Detection
Instrumentation; 3/4.7.5, Fire Suppression Systems; 3/4.7.6, Fire Rated
Assemblies; and 6.1.C.4, Fire Brigade Staffing. The amendments also
replace License Condition 2.C.(25) for Unit 1 and License Condition
2.C.(15) for Unit 2.
Date of issuance: June 10, 1998.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 127 and 112.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the operating licenses and the Technical Specifications.
Date of initial notice in Federal Register: September 25, 1996 (61
FR 50340). The June 19, 1997, and February 2, 1998, supplements
clarified the license conditions by providing specific approval dates
for previous fire protection safety evaluations. This information was
within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: March 13, 1998, as supplemented
March 30, 1998.
Brief description of amendment: The amendment revises the auxiliary
feedwater system technical specification to allow two auxiliary
feedwater flow control valves in one train to be inoperable for up to
72 hours.
Date of issuance: June 10, 1998.
Effective date: June 10, 1998.
Amendment No.: 183.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19967) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423-3698
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: May 20, 1998 (NRC-98-0099)
Description of amendment request: The amendment revises the action
specified in Technical Specification 3.1.3.1, ``Control Rod
Operability,'' by changing the action statements associated with the
scram discharge volume vent and drain valves to align with those in the
NUREG-1433, Revision 1, ``Standard Technical
[[Page 35999]]
Specifications General Electric Plants, BWR/4.''
Date of issuance: June 12, 1998.
Effective date: June 12, 1998.
Amendment No: 120.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
considerations (NSHC): Yes (63 FR 29254 dated May 28, 1998). 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 June 29, 1998, 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 no significant
hazards consideration are contained in a Safety Evaluation dated June
12, 1998.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter.
Duke Energy Corporation, 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: May 22, 1998.
Brief description of amendments: The amendments revise Surveillance
Requirement Section 4.4.3.3 of each unit's Technical Specification to
be consistent with the plant design; specifically, deleting the
reference to manual transfer of power supply from normal to emergency.
Date of issuance: June 17, 1998.
Effective date: As of the date of issuance.
Amendment Nos.: Unit 1--166; Unit 2--158.
Facility Operating License Nos. NPF-35 and NPF-52: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. (63 FR 29759 dated June 1, 1998). That 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 July 1, 1998, but indicated that if the Commission makes a
final no significant hazards consideration determination, any such
hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and final no significant hazards consideration
determination are contained in a Safety Evaluation dated June 17, 1998.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: March 20, 1998.
Brief description of amendment: The amendment revised the Improved
Technical Specification 5.6.2.8 to reflect the current schedule for
performing the required reactor coolant pump flywheel inspection.
Date of issuance: June 8, 1998.
Effective date: June 8, 1998.
Amendment No.: 167.
Facility Operating License No. DPR-31: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 6, 1998 (63 FR
25110).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: November 22, 1996, as revised
and replaced February 2, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) to allow for the installation of a
temporary fuel oil storage and transfer system in order to maintain the
operability of one Unit 3 emergency diesel generator during the
performance of a required surveillance to clean the permanent fuel oil
storage tank.
Date of issuance: June 9, 1998.
Effective date: June 9, 1998.
Amendment Nos.: 197 and 191.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the TS.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9604).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: March 23, 1998.
Description of amendment request: The proposed change would revise
the Seabrook Station Technical Specifications (TSs) to add a new TS
3.0.5 that would provide an exception to TSs 3.0.1 and 3.0.2 to allow
the performance of required testing to demonstrate the operability of
the equipment being returned to service or the operability of other
equipment.
Date of issuance: June 16, 1998.
Effective date: As of its date of issuance, to be implemented
within 60 days.
Amendment No.: 57.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19972)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut.
Date of application for amendment: December 8, 1997.
Brief description of amendment: The changes modify the Technical
Specifications to resolve several compliance issues by rewording of the
text, changing terminology, correcting a
[[Page 36000]]
mode applicability, correcting a formula, updating the Design Features
section, and updating the Bases section to reflect the changes.
Date of issuance: June 16, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 216.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4319).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: February 14, 1997, as
supplemented by letters dated October 9, 1997, March 31, 1998, and
April 15, 1998.
Brief description of amendments: The amendments revised the
combined Technical Specifications (TS) for the Diablo Canyon Power
Plant, Unit Nos. 1 and 2 to change the surveillance frequencies from at
least once every 18 months to at least once per refueling interval
(nominally 24 months) for (1) eight slave relays, (2) 20 electrical
system tests, (3) one electrical Bases change, and (4) five
miscellaneous tests.
Date of issuance: June 5, 1998.
Effective date: June 5, 1998, to be implemented within 90 days from
the date of issuance.
Amendment Nos.: Unit 1--126; Unit 2--124.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14466). The October 9, 1997, March 31, 1998, and April 15, 1998,
supplemental letters provided additional information and did not change
the staff's initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated June 5, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: January 27, 1998.
Brief description of amendments: These amendments revise Table
3.6.3-1 of the Technical Specifications by removing the isolation time
for the high pressure coolant injection turbine exhaust valves and
adding a notation that the isolation is not required.
Date of issuance: June 16, 1998.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment Nos.: 129 and 90.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 11, 1998 (63 FR
11921).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 25, 1997, as supplemented
by letter dated June 2, 1998.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) to allow for up to +17\1/2\ steps of control rod
misalignment when power is greater than 85%.
Date of issuance: June 17, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 180.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 27, 1997 (62 FR
45461)
The June 2, 1998, supplement provided a clarification to the
wording of the TSs and did not change the staff's proposed finding of
no significant hazards consideration. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
June 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: September 3, 1997.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) by revising the number of hours operating
personnel can work in a normal shift. The proposed amendment also
contains some administrative changes to the TS.
Date of issuance: June 17, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 181
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54875).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
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: February 13, 1998 (TS 97-03).
Brief description of amendments: The amendments change the
Technical Specifications by adding a new Limiting Condition for
Operation 3.7.1.6 that addresses the requirements for the main
feedwater isolation valve functions required by the Sequoyah Nuclear
Plant accident analysis.
Date of issuance: June 8, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: Unit 1-232; Unit 2-222.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19979).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 8, 1998.
No significant hazards consideration comments received: No.
[[Page 36001]]
Local Public Document Room location: Chattanooga-Hamilton County
Library, 101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: April 29, 1998.
Brief description of amendment: The requested changes would allow,
temporarily, both trains of hydrogen igniters to be declared inoperable
for up to 72 hours.
Date of issuance: June 9, 1998.
Effective date: June 9, 1998.
Amendment No.: 10.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 7, 1998 (63 FR
25243).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 9, 1998.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: August 26, 1997.
Brief description of amendment: This amendment changed Technical
Specification (TS) Section 3/4.2, ``Power Distribution Limits.'' The
departure from nucleate boiling parameters limiting condition for
operation was modified due to an industry notification.
Date of issuance: June 11, 1998.
Effective date: June 11, 1998.
Amendment No.: 222.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 8, 1997 (62 FR
52590)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: August 26, 1997.
Brief description of amendment: This amendment revises Technical
Specification (TS) Section 3/4.6.1.3, ``Containment Systems--
Containment Air Locks,'' and the associated bases. The limiting
condition for operation and the surveillance requirements were
modified. The application also proposed a change to TS Bases 3/4.9.4,
``Refueling Operations--Containment Penetrations.'' That bases change
was approved by letter dated March 19, 1998.
Date of issuance: June 11, 1998.
Effective date: June 11, 1998.
Amendment No.: 223.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54876)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: December 18, 1997.
Brief Description of amendments: These amendments revise the
Technical Specifications (TS) to clarify the terminology used for
describing equipment surveillances performed on a refueling interval
frequency, and to use consistent wording.
In two cases the proposed changes are denied. These two exceptions,
TS 4.6.A.1.b and 4.6.C.1.e, do not include required specific Mode
restrictions and could not be approved at this time. If appropriate
revisions are submitted, these two exceptions could be found to be
acceptable at a later time.
Date of issuance: June 11, 1998.
Effective date: June 11, 1998.
Amendment Nos.: 213 and 213.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: May 6, 1998 (63 FR
25118). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: November 5, 1997, as
supplemented January 28, 1998 and May 12, 1998.
Brief Description of amendments: These amendments permit an
increase in the maximum allowable fuel enrichment for core reloads from
4.1 to 4.3 weight percent U\235\.
Date of issuance: June 19, 1998.
Effective date: June 19, 1998.
Amendment Nos.: 214 and 214.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68320)
The January 28 and May 12, 1998 submittals provided clarifying
information that did not affect the initial no significant hazards
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 19, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: March 25, 1998.
Brief Description of amendments: These amendments revise the
Technical Specifications to change certain management titles. There is
no change in duties or responsibilities proposed. Specifically, the
Station Manager's title is changed to Site Vice President. The title of
Assistant Station Manager Operations and Maintenance is changed to
Manager-Operations and Maintenance. The title of Assistant Station
Manager Nuclear Safety and Licensing is changed to Manager-Station
Safety and Licensing.
Date of issuance: June 19, 1998.
Effective date: June 19, 1998.
Amendment Nos.: 215 and 215.
[[Page 36002]]
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: May 6, 1998 (63 FR
25119) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 19, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: October 13, 1997, supplemented
on February 10, 1998.
Brief description of amendment: The amendment involves
miscellaneous changes to the TS to (1) relocate information to the
Updated Safety Analysis Report (USAR), (2) delete redundant
information, (3) incorporate new references, (4) delete incorrect
references, (5) correct errors, and (6) augment existing requirements.
Date of issuance: June 9, 1998.
Effective date: June 9, 1998.
Amendment No.: 137.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 8, 1998 (63 FR
11926).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001
Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power
Station, Franklin County, Massachusetts
Date of application for amendment: September 5, 1997 and March 30,
1998.
Brief description of amendment: Revises Technical Specifications
and bases in order to allow loads of up to 80-tons to travel over the
spent fuel pool.
Date of issuance: June 17, 1998.
Effective date: June 17, 1998.
Amendment No.: 149.
Facility Opertating (Possession Only) License No. DPR-3: Amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54879) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301
Dated at Rockville, Maryland, this 24th day of June 1998.
For The Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV Office of Nuclear
Reactor Regulation.
[FR Doc. 98-17352 Filed 6-30-98; 8:45 am]
BILLING CODE 7590-01-P
