[Federal Register Volume 63, Number 116 (Wednesday, June 17, 1998)]
[Notices]
[Pages 33103-33119]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-16012]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving no Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 22,
[[Page 33104]]
1998, through June 5, 1998. The last biweekly notice was published on
June 3, 1998 (63 FR 30261).
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 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 17, 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,
[[Page 33105]]
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.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of application for amendment request: May 18, 1998.
Description of amendment request: Change various technical
specification (TS) values to conservatively reflect design values.
These TS values affect: (1) 125/250 volts direct current (Vdc)
electrolyte temperature; (2) control rod drive accumulator pressure;
(3) standby liquid control solution temperature; (4) ultimate heat sink
minimum water level; (5) shutdown suppression chamber level (Quad
Cities only); and (6) degraded voltage setpoint (Quad Cities 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:
Does the change involve a significant increase in the probability
or consequences of an accident previously evaluated?
The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
proposed changes to certain Technical Specification acceptance values
are conservative and serve to ensure operability of equipment important
to safety. By ensuring equipment availability, the probability or
consequences of an accident previously evaluated are not increased. In
addition, the proposed changes have no impact on any initial condition
assumptions for accident scenarios. Onsite or offsite dose consequences
resulting from an event previously evaluated are not affected by this
proposed amendment request.
Accordingly, there is no significant change in the probability or
consequences of an accident previously evaluated.
Does the change create the possibility of a new or different kind
of accident from any accident previously evaluated?
The proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated. The
proposed license amendment provides changes in certain Technical
Specification values to restore margin and ensure equipment
operability. Each proposed change is conservative with respect to
current requirements. The proposed amendment does not involve any plant
physical changes that would create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, the proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed change does not involve a significant reduction in a
margin of safety. In fact, the proposed changes restore margin and
ensure equipment operability. Since the changes maintain the necessary
level of system reliability, they do not involve a significant
reduction in the margin of safety.
Therefore, the change does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: May 2, 1998, as supplemented May
21, and 23 (three letters), 1998.
Brief description of amendment: This amendment changed Technical
Specification (TS) 3/4.6.2, ``Protective Instrumentation,'' and its
associated Bases to reflect modifications to the initiation
instrumentation for the Control Room Air Treatment System. It also
changed TS 3.2.4a, ``Reactor Coolant Activity,'' and added an
additional condition to the operating license.
Date of issuance: May 23, 1998.
Effective date: As of the date of issuance to be implemented prior
to resumption of power operation.
Amendment No.: 161.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes (63 FR 27601 dated May 19, 1998.
The notice recognized the existence of exigent circumstances pursuant
to 10 CFR 50.91(a)(6) and provided an opportunity to submit comments on
the Commission's proposed no significant hazards consideration
determination. The notice published May 19, 1998, also provided for an
opportunity to request a hearing by June 1, 1998 (this will be
corrected to June 18, 1998, by a notice to be published in the near
future), but indicated that if the Commission makes a final no
significant hazards consideration determination, any such hearing would
take place after issuance of the amendment. Subsequent to publishing
the notice, and due to schedule improvements which have occurred at the
plant, the Commission has determined that the amendment should be
issued on an emergency basis pursuant to 10 CFR 50.91(a)(5). The
Commission's related evaluation of the amendment, finding of emergency
circumstances, consultation with the State of New York, and final no
significant hazards consideration determination are contained in a
Safety Evaluation date May 23, 1998.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State
[[Page 33106]]
University of New York, Oswego, New York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: May 14, 1998.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) for the Reactor Protection
System (RPS) and the Engineered Safety Features Actuation System
(ESFAS) instrumentation by restricting the time most RPS and ESFAS
actuation channels can be in the bypass position to 48 hours. The
current TSs have no time limit. The proposed amendment would also
modify the TS action requirements and the channel calibration
requirements for the loss of turbine load reactor trip function, and
the channel calibration requirements for the wide range logarithmic
neutron flux monitors; add a note to exclude the neutron detectors from
the channel calibration requirements; correct a reference to a TS
surveillance requirement; and correct errors that have been identified.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to restrict the time most of the reactor
protection or engineered safety feature actuation channels can be in
the bypass position to 48 hours, from an indefinite period of time, has
no effect on the design of the Reactor Protection System (RPS) or the
Engineered Safety Feature Actuation System (ESFAS), and does not affect
how these systems operate. In addition, this will minimize the
susceptibility of these systems to the remote possibility of fault
propagation between channels. The pressurizer high pressure reactor
protection channels will not be required to be placed in the tripped
condition after 48 hours. A failed pressurizer high pressure channel
will be allowed to remain in the bypassed condition for up to 30 days.
If the failed pressurizer high pressure channel was placed in the
tripped condition, and then a high failure of another pressurizer high
pressure channel occurred, the reactor would trip and both pressurizer
power operated relief valves (PORVs) would open, resulting in an
undesired loss of primary coolant. Limiting the time that a failed
pressurizer high pressure reactor protection channel can be in bypass
to 30 days will minimize the risk of the inadvertent opening of both
PORVs, as well as the risk associated with fault propagation between
channels. These systems will still function as designed to mitigate
design basis accidents. Therefore, this change does not significantly
increase the probability or consequences of an accident previously
evaluated.
The proposed change to increase the time a second RPS or ESFAS
channel can be removed from service (from 2 hours to 48 hours),
provided one of the inoperable channels is placed in the tripped
condition, has no effect on the design of the RPS or ESFAS and does not
affect how these systems operate. These systems will still function as
designed to mitigate design basis accidents.
However, one of the proposed changes will allow two pressurizer
pressure reactor protection channels to be removed from service (one
channel in the tripped condition and one channel in the bypassed
condition) for 48 hours instead of the current 2 hour time limit. With
a pressurizer pressure channel in the tripped condition, the high
failure of a second pressurizer pressure channel would initiate a
reactor trip, open both pressurizer PORVs, and cause an undesired loss
of primary coolant. Thus, this change will increase the probability of
occurrence of a previously evaluated accident (FSAR [Final Safety
Analysis Report] Section 14.6.1--Inadvertent Opening of a Pressurized
Water Reactor Pressurizer Pressure Relief Valve). However, since this
configuration will only be allowed for an additional 46 hours, the
increase in the probability of occurrence of a previously evaluated
accident will be limited to an acceptable value. Therefore, this change
does not significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change to apply a more restrictive action statement to
the loss of turbine load reactor trip function has no effect on the
design of this trip function and does not affect how this trip function
operates. Also, this trip function is not assumed to operate to
mitigate any design basis accident.
Therefore, this change does not significantly increase the
probability or consequences of accident previously evaluated.
The proposed change to require a channel calibration every 18
months for the loss of turbine load reactor trip function and for the
wide range logarithmic neutron flux monitors has no effect on the
design of either the loss of turbine load reactor trip function or the
wide range logarithmic neutron flux monitors. Also, neither of these
are assumed to operate to mitigate any design basis accident.
Therefore, this change does not significantly increase the probability
or consequences of an accident previously evaluated.
The proposed change to exclude the neutron detectors from the
channel calibration requirement has no effect on the design of the
neutron detectors and has no significant effect on how these detectors
operate. The detectors are passive devices with minimal drift. In
addition, slow changes in the sensitivity of the linear power range
flux detectors is compensated for by performing the daily calorimetric
calibration and the monthly calibration using the incore detectors.
These detectors will still function as designed to mitigate design
basis accidents. Therefore, this change does not significantly increase
the probability or consequences of an accident previously evaluated.
The proposed change to correct the surveillance requirement
referenced in an action statement has no effect on the design of the
ESFAS and does not affect how this system operates. The ESFAS will
still function as designed to mitigate design basis accidents.
Therefore, this change does not significantly increase the probability
or consequences of an accident previously evaluated.
The proposed change to add a reference to the reactor coolant pump
low speed reactor trip function to a note that states this trip may be
bypassed when [less than] 5 [percent] power, and that the bypass must
be automatically removed when [greater than or equal to] 5 [percent]
power will not effect this reactor trip function. This bypass
capability currently exists in the design of the Millstone Unit No. 2
RPS, and is the same bypass feature referenced for the reactor coolant
flow low reactor trip function. Both of these reactor trip functions
provide protection for a reduction in RCS [Reactor Coolant System]
flow. The addition of this note will not result in any technical change
to the Millstone Unit No. 2 RPS. The RPS will continue to function as
before. Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
[[Page 33107]]
The proposed change to correct the power level high trip setpoint
on Technical Specification Page 2-4 will not result in any change to
the actual plant setpoint for this RPS trip function. As a result of
this proposed change, the setpoint listed on Page 2-4 will agree with
the setpoint previously approved by the NRC, and currently used by the
RPS. The change has no effect on the design of the RPS and does not
affect how this system operates. Therefore, this change does not
significantly increase the probability or consequences of an accident
previously evaluated.
The information added to the Bases of the Technical Specifications
to provide a discussion of how the RPS and ESFAS are affected by the
proposed changes, the effect the action statements have on the
operation of the RPS and ESFAS, and to discuss the impact of
surveillance testing on RPS operability will have no effect on
equipment operation. The RPS and ESFAS will continue to function as
designed to mitigate design basis accidents. Therefore, this change
does not significantly increase the probability or consequences of an
accident previously evaluated.
Thus, this License Amendment Request does not impact the
probability of an accident previously evaluated nor does it involve a
significant increase in the consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no new
or different type of equipment will be installed) or require any new or
unusual operator actions. They do not alter the way any structure,
system, or component functions and do not alter the manner in which the
plant is operated. The proposed changes do not introduce any new
failure modes. They will not alter assumptions made in the safety
analysis and licensing basis. The RPS and the ESFAS will still function
as designed to mitigate design basis accidents.
Therefore, these changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any safety analysis assumption. The proposed
changes do not decrease the scope of equipment currently required to be
operable or subject to surveillance testing, nor do the proposed
changes affect any instrument setpoints or equipment safety functions.
The effectiveness of Technical Specifications will be maintained
since the changes will not alter the operation of any RPS or ESFAS
function. In addition, most of the changes are consistent with the
Calvert Cliffs RPS and ESFAS Technical Specifications mode provided in
Enclosure 3 of the NRC correspondence dated April 16, 1981 (R. A. Clark
letter to W. G. Counsil, Evaluation of the Reactor Protection System
Inoperable Channel Condition at Millstone Nuclear Power Station, Unit
No. 2, dated April 16, 1981) and the new, improved Standard Technical
Specifications (STS) for Combustion Engineering plants (NUREG-1432).
Therefore, there is no 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: 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.
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 amendment request: June 25, 1997.
Description of amendment request: The proposed amendment would
change the Indian Point 3 Technical Specifications to allow the use of
zirconium alloy or stainless steel filler rods in fuel assemblies to
replace failed or damaged fuel rods.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no significant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident previously
analyzed?
Response: The proposed changes modify the technical specification
only to the extent that the reconstitution is recognized as acceptable
under limited circumstances. Reconstitution is limited to substitution
of zirconium alloy or stainless steel filler rods, and must be in
accordance with approved applications of fuel rod configurations.
Although these changes permit reconstitution to occur without the need
for a specific technical specification change, use of an approved
methodology is required prior to its application. Since the changes
will allow substitution of filler rods for leaking, potentially leaking
rods or damaged rods, the changes may actually reduce the radiological
consequences of an accident. It is noted that the specific changes
requested in this letter have previously been found acceptable by the
NRC in GL [Generic Letter] 90-02, Supplement 1. For these reasons, we
conclude that the changes will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
(2) Does the proposed license amendment create the possibility of a
new or different kind of accident from any previously evaluated?
Response: The proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated because they will only affect the assembly configuration and
can only be implemented if demonstrated to meet current plant
requirements in accordance with an NRC-approved methodology. The other
aspects of plant design, operation limitations, and responses to events
will remain unchanged. It is noted that the changes have previously
been determined acceptable by the NRC in GL 90-02, Supplement 1.
(3) Does the proposed amendment involve a significant reduction in
a margin of safety?
Response: The proposed change will not involve a reduction in a
margin of safety because the changes can only be implemented if
demonstrated to meet current plant requirements in accordance with an
NRC-approved methodology. It is noted that the changes have previously
been determined acceptable by the NRC in GL 90-02, Supplement 1.
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.
[[Page 33108]]
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: April 28, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.2.1 to replace the plus or
minus 1 percent setpoint tolerance limit for safety/relief valves
(SRVs) with a plus or minus 3 percent setpoint tolerance limit. In
addition, the proposed amendment would revise TS 4.4.2.2 to state that
all SRVs must be certified to be within plus or minus 1 percent of the
TS setpoint prior to returning the valves to service.
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 involve: (1) no significant hardware
changes; (2) no significant changes to the operation of any systems or
components in normal or accident operating conditions; and (3) no
changes to existing structures, systems, or components. Therefore these
changes will not increase the probability of an accident previously
evaluated.
These proposed changes were developed in accordance with the
provisions contained in an NRC Safety Evaluation Report, dated 3/8/93,
for the ``BWR Owners Group Inservice Pressure Relief Technical
Specification [Revision] Licensing Topical Report'', NEDC-31753P as
described in General Electric report NEDC-32511P, ``Safety Review for
Hope Creek [Generating Station] Safety/Relief Valve Tolerance
Analyses''. Since the plant systems associated with these proposed
changes will still be capable of: (1) meeting all applicable design
basis requirements; and (2) retain the capability to mitigate the
consequences of accidents described in the HC [Hope Creek] UFSAR
[Updated Final Safety Analysis Report], the proposed changes were
determined to be justified. Therefore, these changes will not involve a
significant increase in the consequences of an accident previously
evaluated.
(2) The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Establishment of the [plus or minus] 3 [percent] SRV setpoint
tolerance limit will not adversely impact the operation of any safety
related component or equipment. Since the proposed changes involve: (1)
no significant hardware changes; (2) no significant changes to the
operation of any systems or components; and (3) no changes to existing
structures, systems, or components, there can be no impact on the
occurrence of any accident. These proposed changes were developed in
accordance with the provisions contained in an NRC Safety Evaluation
Report, dated 3/8/93, for the ``BWR Owners Group Inservice Pressure
Relief Technical Specification [Revision] Licensing Topical Report'',
NEDC-31753P as described in General Electric report NEDC-32511P,
``[Safety Review for Hope Creek Generating Station] Safety/Relief Valve
Tolerance Analyses''. Furthermore, there is no change in plant testing
proposed in this change request which could initiate an event.
Therefore, these changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) The proposed change does not involve a significant reduction in
a margin of safety.
Establishment of the [plus or minus] 3 [percent] SRV setpoint
tolerance limit will not adversely impact the operation of any safety
related component or equipment. General Electric analyses performed for
Hope Creek and contained in General Electric report NEDC-32511P,
``[Safety Review for Hope Creek Generating Station] Safety/Relief Valve
Tolerance Analyses,'' concluded that there is no significant impact on
fuel thermal limits, no significant impact on safety related systems,
structures or components, and no significant impact on the accident
analyses associated with the proposed changes. Therefore, the changes
contained in this request do not result in a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, 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.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric
Generating Plant, Units 1 and 2, Burke County, Georgia
Date of amendment request: May 8, 1998.
Description of amendment request: The proposed amendments would
change the Vogtle Electric Generating Plant (VEGP) Technical
Specification (TS) 5.5.7, ``Reactor Coolant Pump Flywheel Inspection
Program,'' to provide an exception to the examination requirements of
Regulatory Position C.4.b of Regulatory Guide (RG) 1.14, Revision 1,
August 1975.
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 safety function of the RCP [reactor coolant pump] flywheel is
to provide sufficient rotational inertia to ensure reactor coolant flow
through the core during coastdown following a loss of offsite power and
subsequent reactor trip. FSAR [Final Safety Analysis Report] Chapter 15
analysis for a complete loss of forced reactor coolant flow
demonstrates that the reactor trip together with the flow sustained by
the inertia of the RCP impeller will be sufficient to prevent the most
limiting fuel assembly from exceeding the DNBR [departure from nucleate
boiling ratio] limits.
The maximum mechanical loading on the RCP motor flywheel results
from overspeed following a LOCA [loss-of-coolant accident]. The
analysis presented in WCAP-14535A demonstrates that the revised
inspection program proposed by this license amendment will ensure the
integrity of the RCP flywheels will be maintained.
[[Page 33109]]
Based upon the findings of WCAP-14535A, the ability of the RCP
flywheel to perform its intended safety function will be unaffected by
the license amendment and the FSAR Chapter 15 analysis will remain
valid. Therefore, these proposed changes do 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 amendment will not change the physical plant
configuration nor the modes of operation of any plant equipment. Based
upon the results of WCAP-14535A, no new failure mechanism will be
introduced by the revised RCP flywheel inspection program. Therefore,
the proposed amendment will 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 operating limits and functional capabilities of the affected
systems, structures, and components will be unchanged by the proposed
amendment. The results of the RCP flywheel inspections performed
throughout the industry and at VEGP have identified no indications
which would affect its integrity. As presented in WCAP-14535A, detailed
stress analysis and risk assessments have been completed with the
results indicating that there would be no change in the probability of
failure for RCP flywheels if all inspections were eliminated.
Therefore, these changes do not involve a significant reduction in 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: May 6, 1998.
Description of amendment request: The proposed amendment would
replace the two percent penalty addressed in surveillance requirement
(SR) 3.2.1.2(a) with a burnup-dependent factor to be specified in the
Watts Bar Core Operating Limits Report (COLR). Specifically, the
following changes are being proposed:
1. SR 3.2.1.2(a) and its associated BASES will have the phrase ``by
a factor of 1.02'' deleted and replaced with the phrase ``by the
appropriate factor specified in the COLR.''
2. Technical Specification (TS) Section 5.9.5(b)(3) would be
updated to reference the revised WCAP (10216-P-A, Revision 1A, 1994)
that details the analytical methods utilized for the new penalty
factor.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The proposed change involves only the manner in which the penalty
factors for FQ(Z) would be specified (i.e., burnup-dependent
factor specified in the Core Operating Limits Report [COLR] versus a
constant factor specified in the TS). This is simply used to account
for the fact that FQ C(Z) may increase between
surveillance intervals. These penalty factors are not assumed in any of
the initiating events for the accident analyses. Therefore the proposed
change will have no effect on the probability of any accidents
previously evaluated. The penalty factors specified in the COLR will be
calculated using NRC-approved methodology and will continue to provide
an equivalent level of protection as the existing TS requirement.
Therefore, the proposed change will not affect the consequences of any
accident previously evaluated.
B. The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed change does not involve a physical alteration to the
plant (no new or different kind of equipment will be installed) or
alter the manner in which the plant would be operated. Thus, this
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
C. The proposed amendment does not involve a significant reduction
in a margin of safety.
The proposed change will continue to ensure that potential
increases in FQ C(Z) over a surveillance interval
will be properly accounted for. The penalty factors will be calculated
using an NRC-approved methodology. Therefore, the proposed change will
not involve a reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: May 1, 1998.
Description of amendment request: The proposed amendment would make
several editorial changes to the Administrative Controls section of the
Technical Specifications. The changes include revisions due to
organizational changes, quality assurance changes, editorial changes,
and typographical corrections.
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 proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The administrative change proposed herein will have no effect on
plant hardware, plant design, safety limit setting or plant system
operation and therefore do[es] not modify or add any initiating
parameters that would significantly increase the probability or
consequences of any previously analyzed accident. The proposed
amendment changes the reference to the VYNPS QA program and makes other
[[Page 33110]]
administrative changes, such as title changes and correction/
clarification of errors. Therefore, there is no increase in the
probability or consequence of an accident previously evaluated.
2. Will the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
This change does not affect any equipment nor does it involve any
potential initiating events that would create any new or different kind
of accident. The proposed change involves [ ] wording changes in the
Technical Specifications identifying the name of the QA program and
makes other administrative changes, such as title changes and
corrective/clarification of errors. Therefore no new or different kind
of accident has been introduced.
3. Will the proposed changes involve a significant reduction in a
margin of safety?
This change does not affect any equipment involved in potential
initiating events or safety limits. The proposed change has no
significant impact on margin of safety, as it is comprised of only
administrative changes.
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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Project Director: Cecil O.Thomas.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: September 1, 1995, as supplemented April
8, 1996, April 22, 1996, April 23, 1996, November 18, 1997, February 9,
1998, March 25, 1998 and May 5, 1998. This notice supersedes the
Federal Register notice of September 27, 1995 (60 FR 49949)
Description of amendment request: The originally (September 1,
1995) proposed changes to the Technical Specifications (TS) would
permit a single outage of up to 14 days for each emergency diesel
generator (EDG) once every 18 months in order to perform preventive
maintenance. The amended request will permit a single outage of up to
14 days for each EDG for any reason; TS change to incorporate a
Configuration Risk Management Program (CRMP) in the Administrative
Section in the TS, in support of the previous submittal for the 14-day
Allowed Outage Time (AOT) for the EDGs and would permit an increase in
the TS maintenance interval of the EDG from 18 to 24 months, based on
the recommendation from the EDG owners group (Fairbanks Morse Owners
Group).
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. Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specification changes will not:
a. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
A probabilistic safety analysis (PSA) has been performed which
demonstrates that a 14-day AOT for each EDG, results in a small change
in core damage frequency assuming adequate compensatory measures are in
place. The compensatory measures include requirements that the other
EDGs, off-site power supply, and the alternate A.C. diesel (AAC DG) be
operable whenever the action statement is entered.
The effect of the proposed change has been calculated to be an
increase in core damage frequency of approximately 1 E-6 per year from
the baseline core damage frequency of 4.1 E-5. Considering that credit
was not taken for the AAC DG previously in the IPE nor was the AAC DG
specified in Technical Specifications, the proposed changes remain
bounded by the core damage frequency identified in the Individual Plant
Examination.
Credit for the AAC DG was previously not taken nor was the AAC DG
previously included in the Technical Specifications. Furthermore, the
probabilistic safety analysis (PSA) demonstrates that the increase in
core damage frequency due to extending the EDG AOT of a 14-day period
is not significant as long as the AAC DG is operable to act as a source
of emergency power to replace the EDG. The period of time during which
the EDG is unavailable is short enough to limit the impact of using the
manually operated AAC DG as a replacement for the automatically
operated EDG.
The plant design and operation are not changed by the incorporation
of a CRMP into the Administrative Section of Technical Specifications.
Further, with the proposed change to the preventive maintenance
interval, the EDG reliability remains adequate to perform its function
of supporting accident mitigation equipment with emergency electrical
power.
Therefore, neither the probability of occurrence nor the
consequences of an accident or malfunction of equipment important to
safety previously evaluated in the safety analysis report are increased
due [to] the proposed changes to permit a 14-day allowed outage time
and a 24 month preventive maintenance interval for the EDGs.
b. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
No new initiators are defined as a result of a review of the PSA
model. The proposed Technical Specifications changes only modify the
AOT of an EDG. The UFSAR [Updated Final Safety Analysis Report]
accidents are analyzed assuming that the EDG is the worst single
failure. This assumption is more severe than the proposed Technical
Specifications changes, which [replace] the EDG with the AAC DG.
Similarly, the PSA performed to evaluate the proposed Technical
Specifications changes considered all of the initiating events defined
for the PSA performed for the Individual Plant Examination. No new
initiators were defined as a result of a review of the PSA model.
Adding the CRMP and changing the EDG preventive maintenance
interval in the Technical Specifications does not change any method of
operation or create any new modes of operation or accident precursors.
Therefore, it is concluded that no new or different kind of
accident or malfunction from any previously evaluated has been or will
be created by the proposed changes to permit a 14-day allowed outage
time and a 24 month preventive maintenance interval for the EDGs.
c. The proposed Technical Specifications changes do not result in a
reduction in margin of safety as defined in the basis for any Technical
Specifications.
The PSA was performed to evaluate the concept of a one-time outage.
The results of the analyses show a small change in the core damage
frequency. As described above the proposed Technical Specifications
changes only modify the AOT of an EDG. Thus, operation with slightly
increased EDG unavailability due to maintenance is acceptable given the
operability of the AAC DG and the other EDG.
Incorporating the CRMP and changing the EDG preventive maintenance
interval in the Technical Specifications
[[Page 33111]]
does not affect any accident analysis assumptions or change any
Technical Specifications criteria.
Therefore, the margin of safety is not changed.
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: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Gordon E. Edison, Acting.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: May 7, 1998.
Description of amendment request: Technical Specification 5.4,
``Fuel Storage,'' would be changed to increase the allowable mass of
uranium-235, per axial centimeter, for fuel storage in new fuel and
spent fuel storage racks. This change will allow use of new Siemens
heavy fuel assemblies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
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 mass of the fuel assembly is increased by a small amount (30
pounds, or 2.4%), from that of the fuel assemblies now in the core.
Even with this increase, the load on the fuel handling equipment is
still well within design limits. Therefore, the probabilities of a fuel
handling accident inside containment (FHAIC) and the fuel handling
accident outside containment (FHAOC) are not changed.
The total core mass, with Siemens heavy fuel, is less than that
assumed in the original plant safety analysis. The proposed change does
not alter the plant configuration, operating set points, or overall
plant performance. The probability of other accidents is therefore not
changed.
Attachment 4 (of the application) shows that the consequences of a
fuel handling accident or a large break loss of coolant accident are
not significantly affected.
Any changes in the nuclear properties of the reactor core that may
result from a higher mass of fuel U235 per axial centimeter
will be analyzed and shown to meet acceptance criteria in the
appropriate reload analysis, which would be completed prior to use.
(2) Create the possibility of a new or different kind of accident
from any previously evaluated.
As discussed above, the only safety issue significantly affected by
the proposed change is the criticality analysis of the spent fuel
storage racks and new fuel storage racks. Since it has been
demonstrated that keff remains below the keff
acceptance criteria, no new or different accident would be created
through the use of fuel with up to 56.067 grams of U235 per
axial centimeter at the Kewaunee Nuclear Power Plant.
The proposed change does not alter the plant configuration,
operating set points, or overall plant performance and therefore does
not create a new or different kind of accident from any accident
previously evaluated.
(3) Involve a significant reduction in the margin of safety.
The criticality analysis in Reference 3 (of the application)
demonstrates that adequate margins to criticality can be maintained
with up to 56.067 grams of U235 per axial centimeter stored
in either the new fuel storage racks or the spent fuel storage racks.
The bounding cases of the analysis demonstrate that keff
remains less than 0.95 in the spent fuel storage racks and the new fuel
storage racks if flooded with unborated water. The bounding cases of
the analysis also demonstrate that keff remains less than
0.98 in the new fuel storage racks if moderated by optimally misted
moderator. Therefore, the 56.067 grams of U235 per axial
centimeter limit is acceptable for storage in both the new fuel storage
racks and the spent fuel storage racks.
Any changes in the nuclear properties of the reactor core that may
result from a higher mass of fuel U235 per axial centimeter
will be analyzed in the appropriate reload analysis to ensure
compliance with applicable reload considerations and requirements.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
Wisconsin Electric Power Company, Docket No. 50-301, Point Beach
Nuclear Plant, Unit 2, Town of Two Creeks, Manitowoc County, Wisconsin
Date of amendment request: May 15, 1998 (NPL-98-0303).
Description of amendment request: The proposed amendment revises
the schedule for implementing the boron concentration changes related
to the planned conversion of Unit 2 to 18-month fuel cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendment will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes are administrative only. There are no physical
changes to the facility or its operation. All Limiting Conditions of
Operation, Limiting Safety System Settings, and Safety Limits specified
in the Technical Specification remain unchanged. Additionally, there
are no changes in the Quality Assurance Program, Emergency Plan,
Security Plan, and Operator Training and Requalification Program.
Therefore, an increase in the probability or consequences of an
accident previously evaluated cannot occur.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes are administrative only. No changes to the
facility structures, systems and components or their operation will
result. The design and design basis of the facility remain unchanged.
The plant safety analyses remain current and accurate. No new or
different failure mechanisms are introduced. Therefore,
[[Page 33112]]
the possibility of a new or different kind of accident from any
accident previously evaluated is not introduced.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendment does not involve a significant reduction in a
margin of safety.
The proposed [amendment is] administrative only. All safety margins
established through the design and facility license including the
Technical Specifications remain unchanged. Therefore, all margins of
safety are maintained.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
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.
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 proposed amendment would
modify the scram discharge volume (SDV) vent and drain valve action
requirements to be consistent with those contained in NUREG-1433,
Revision 1, ``Standard Technical Specifications General Electric
Plants, BWR/4.''
Detroit Edison is requesting that this license amendment request be
processed in an exigent manner in accordance with 10 CFR 50.91(a)(6)
because delay in granting this amendment could lead to a plant
shutdown.
Date of publication of individual notice in Federal Register: May
28, 1998 (63 FR 29254).
Expiration date of individual notice: Comments: June 11, 1998;
hearing: June 29, 1998.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Duke Energy Corporation, Docket Nos. 50-413 and 50-414, Catawba Nuclear
Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 22, 1998.
Description of amendment request: The proposed amendments would
revise Surveillance Requirement Section 4.4.3.3 of the Technical
Specifications. Section 4.4.3.3 currently requires that the emergency
power supply for the pressurizer heaters be demonstrated OPERABLE at
least once per 18 months by manually transferring power from the normal
to the emergency power supply. The licensee proposed to delete the
``manual'' requirement because the power supply transfer at the unit
was designed to be automatic. The proposed requirement is to verify
that required pressurizer heaters are capable of being powered from an
emergency power supply once per 18 months.
Date of publication of individual notice in Federal Register: June
1, 1998 (63 FR 29759).
Expiration date of individual notice: July 1, 1998.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: May 2, 1998.
Brief description of amendment: The amendment changes the Technical
Specifications 3/4.6.2, ``Protective Instrumentation,'' to reflect
modifications to the initiation instrumentation for the Control Room
Air Treatment system.
Date of publication of individual notice in Federal Register: May
19, 1998 (63 FR 27601).
Expiration date of individual notice: June 18, 1998.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: May 15, 1998 (two letters).
Brief description of amendment: The amendment changes
administrative sections of the Technical Specifications to reflect a
restructuring of upper management organization.
Date of publication of individual notice in Federal Register: June
2, 1998 (63 FR 30026).
Expiration date of individual notice: July 2, 1998.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
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: May 12, 1998.
Brief description of amendment request: These amendments relocate
certain requirements related to fire protection from the TSs to the
Updated Final Safety Analysis Report. The TS sections to be relocated
are: 3/4.3.7.9, Fire Detection Instrumentation; 3/4.7.6, Fire
Suppression Systems; 3/4.7.7, Fire Rated Assemblies; and 6.2.2e, Fire
Brigade Staffing. The amendments also replace License Condition 2.C.(6)
for Unit 1 and License Condition 2.C.(3) for Unit 2. These amendments
are consistent with the guidance of NRC Generic Letter (GL) 86-10,
``Implementation of Fire Protection Requirements,'' and GL 88-12,
``Removal of Fire Protection Requirements from Technical
Specifications.''
Date of publication of individual notice in Federal Register: May
21, 1998 (63 FR 28010).
Expiration date of individual notice: June 22, 1998.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
[[Page 33113]]
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: March 31, 1997, as supplemented June 18,
1997, October 10, 1997, October 20, 1997, November 11, 1997, December
22, 1997, January 15, 1998, January 27, 1998, March 30, 1998, April 23,
1998, and April 27, 1998.
Brief description of amendment request: The proposed amendment
would revise the Ginna Station Improved Technical Specifications to
reflect a planned modification to the spent fuel pool storage racks.
Date of publication of individual notice in Federal Register: May
12, 1998 (63 FR 26213). This notice supersedes the March 31, 1997,
application published on April 30, 1997 (62 FR 23502).
Expiration date of individual notice: June 11, 1998.
Local Public Document Room Location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of application for amendments: January 31, 1997, as
supplemented February 13, February 28, March 25, April 16, August 19,
and September 29, 1997, January 22, March 17, April 8, April 21, 1998,
and May 22, 1998.
Brief description of amendments: The amendments revise the TS for a
reduction of the total reactor coolant system flow limit from 370,000
gallons per minute (gpm) to 340,000 gpm in support of increased steam
generator tube plugging.
Date of issuance: May 23, 1998.
Effective date: As of the date of issuance Unit 1 to be implemented
within 60 days and Unit 2 prior to startup from the spring 1999
refueling outage.
Amendment Nos.: 228 and 202.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 26, 1997 (62
FR 8780).
The February 13, February 28, March 25, April 16, August 16, and
September 29, 1997, January 22, March 17, April 8, and April 21, 1998,
and May 22, 1998, letters provided clarifying information that did not
change the initial proposed no significant hazards consideration.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated May 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 18, 1997.
Brief description of amendments: The amendments revise the listed
design suppression chamber temperature of 200 deg.F to 220 deg.F and
the listed total water and steam volume of the reactor coolant system
from 18,670 cubic feet to 18,320 cubic feet, respectively.
Date of issuance: May 27, 1998.
Effective date: May 27, 1998.
Amendment Nos.: 195 and 225.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revise the facility's Technical Specifications.
Date of initial notice in Federal Register: August 27, 1997 (62 FR
45454).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 27, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: June 12, 1997, as supplemented
February 2, 1998. The February 2, 1998, submittal contained clarifying
information only and did not change the initial proposed no significant
hazards consideration or expand the scope of the original Federal
Register Notice.
Brief Description of amendments: The amendments consist of changes
to the Technical Specifications (TS) to revise the Limiting Condition
for Operation of the TS to limit the drywell average air temperature
rather than primary containment air temperature. Additionally, the
amendments require that the drywell average air temperature be
maintained less than or equal to 150 deg.F during plant operation. The
current primary containment average temperature limit is 135 deg.F.
Date of issuance: May 28, 1998.
Effective date: May 28, 1998.
Amendment Nos.: 196 and 226.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: August 27, 1997 (62 FR
45454) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 28, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
[[Page 33114]]
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: October 28, 1997
Brief Description of amendments: The amendments revise certain
instrumentation allowable values in the current technical
specifications to the Improved Technical Specifications format.
Date of issuance: May 28, 1998.
Effective date: May 28, 1998.
Amendment Nos.: 197 and 227. Facility Operating License Nos. DPR-71
and DPR-62: Amendments change the Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68304)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 28, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket Nos. 50-325 & 50-324,
Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County, North
Carolina
Date of amendment request: November 15, 1995.
Brief description of amendment: The amendments modify the channel
functional test interval in the Technical Specifications Surveillance
Requirements for the Electrical Protective Assemblies in the Reactor
Protection System.
Date of issuance: May 29, 1998.
Effective date: May 29, 1998.
Amendment No.: 198 and 228.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34887).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 29, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket Nos. 50-325 & 50-324,
Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County, North
Carolina
Date of amendment request: November 16, 1994, as supplemented by
letters dated February 14, 1995, and April 9, 1998.
Brief description of amendment: The amendments change the Technical
Specifications (TS) for Units 1 and 2 to revise the basis for removing
the suppression chamber water temperature monitoring instrumentation
requirements from the TS. This change is being processed in parallel
with the Improved Technical Specification conversion.
Date of issuance: May 29, 1998.
Effective date: May 29, 1998.
Amendment Nos.: 199 and 229.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
497)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 29, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: April 4, 1996, as supplemented
January 24, 1997, March 31, 1997, April 2, 1997, April 14, 1997, March
24, 1998, and May 20, 1998.
Brief Description of amendments: The amendments modify Technical
Specifications (TS) 3.0.4, 4.0.3, and 4.0.4, and their associated Bases
in accordance with the guidance provided in Generic Letter 87-09,
``Sections 3.0 and 4.0 of the Standard Technical Specifications (STS)
on the Applicability of Limiting Conditions for Operation and
Surveillance Requirements.''
Date of issuance: June 2, 1998.
Effective date: June 2, 1998.
Amendment Nos.: 200 and 230.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: July 17, 1996 (61 FR
37297).
The supplemental submittals contained clarifying information only,
and did not change the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Carolina Power & Light Company, et al., Docket Nos. 50-325 & 50-324,
Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County, North
Carolina
Date of amendment request: April 30, 1997, as supplemented October
28, 1997, and May 15, 1998.
Brief description of amendment: The amendments revise surveillance
requirements 4.7.2.b.2 and 4.7.2.c to require testing of the control
room emergency ventiliation system charcoal adsorber in accordance with
the American Society for Testing and Material D3803-1989, ``Standard
Test Method for Nuclear-Grade Activated Carbon.''
Date of issuance: June 2, 1998.
Effective date: June 2, 1998.
Amendment Nos.: 201 and 231.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: July 30, 1997 (62 FR
40846).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: April 3, 1998
Brief description of amendments: The amendments revise the
specified total volume of the condensate storage tank capacity
requirements from 150,000 gallons to 228,200 gallons to ensure the Core
Spray System requirement of 50,000 gallons.
Date of issuance: June 5, 1998.
Effective date: June 5, 1998.
Amendment Nos.: 202 and 232.
[[Page 33115]]
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revise the facility's Technical Specifications.
Date of initial notice in Federal Register: May 6, 1998 (63 FR
25103).
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: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: October 29, 1997.
Brief description of amendment: This amendment changes Technical
Specifications (TS) 3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 by
eliminating the plant shutdown requirements in these TS, and allowing
the applicable redundant feature TS to direct the plant shutdown when
required.
Date of issuance: May 22, 1998.
Effective date: May 22, 1998.
Amendment No.: 78.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68305).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 22, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: April 24, 1998, as supplemented
by letter dated May 15, 1998.
Brief description of amendment: This amendment revises TS 3.3.2,
``Engineered Safety Features Actuation System Instrumentation,'' such
that surveillance of the undervoltage relays may be performed without
entry into TS 3.0.3. Specifically, the change modifies Table 3.3-3 to
allow operation with more than one channel of the emergency bus
undervoltage relays inoperable.
Date of issuance: June 3, 1998.
Effective date: June 3, 1998.
Amendment No.: 79.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 4, 1998 (63 FR
24574).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 3, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN
50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: September 24, 1997.
Brief description of amendments: The amendments revise the
surveillance frequency for the turbine throttle valves and the turbine
governor valves from monthly to quarterly.
Date of issuance: May 26, 1998.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 103 and 93.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 11, 1998 (63 FR
11917).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 26, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: January 28, 1998 (NRC-98-0006),
as supplemented on March 10, 1998 (NRC-98-0036).
Brief description of amendment: The amendment revises technical
specification surveillance requirement 4.4.3.2.2.a for the leak rate
test of the pressure isolation valves, extending it from the current
18-month interval to a 24-month interval.
Date of issuance: May 28, 1998.
Effective date: May 28, 1998, with full implementation within 90
days.
Amendment No.: 118.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9598).
The March 10, 1998, supplement requested a change in the
implementation period. This information was within the scope of the
original Federal Register notice and did not change the staff's initial
proposed no significant hazards considerations determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated May 28, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: November 22, 1995 (NRC-95-0124),
as supplemented February 19, April 19, May 3, June 12, and December 4,
1996, January 30 and August 7, 1997, and April 27 and May 22, 1998.
Brief description of amendment: The amendment revises technical
specification (TS) 3.8.1.1 to change the emergency diesel generator
(EDG) allowed outage time from 3 to 7 days and add a requirement to
verify that combustion turbine-generator 11-1 is available prior to
removing an EDG from service. In addition, in accordance with draft
staff guidance for risk-informed amendments, a section is added to the
Administrative Controls Section of the TS describing the licensee's
configuration risk management program. The associated Bases are also
revised. The November 22, 1995, submittal also requested changes to the
testing and reporting requirements for the EDGs. These aspects were
addressed in Amendment No. 107 to the TS issued on June 20, 1996. The
staff's action on the licensee's request is now complete.
Date of issuance: June 2, 1998.
Effective date: June 2, 1998, with full implementation within 60
days.
Amendment No.: 119.
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7550) with a supplemental notice on May 1, 1998 (63 FR 24195).
[[Page 33116]]
The February 19, April 19, May 3, June 12, and December 4, 1996,
August 7, 1997, and May 22, 1998, submittals provided clarifying
information within the scope of the Federal Register notices and did
not change the staff's initial proposed no significant hazards
considerations determinations.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: March 17, 1998, as supplemented
May 14, 1998.
Brief description of amendments: These amendments revise Action 34
of technical specification (TS) Table 3.3-3, ``Engineered Safety
Feature Actuation System Instrumentation.'' Action 34 is applicable to
Functional Units 6.b., ``Grid Degraded Voltage (4.16 kV Bus),'' and
6.c., ``Grid Degraded Voltage (480 v Bus).'' Revised Action 34 requires
that with one degraded grid voltage monitoring channel inoperable, the
inoperable channel be placed in the tripped condition within one hour;
otherwise, immediately enter the applicable action statement(s) for the
associated emergency diesel generator made inoperable by the degraded
voltage start instrumentation. The revision to Action 34 also requires
that with two degraded grid voltage monitoring channels inoperable,
within one hour restore at least one of the channels to operable status
and place the other channel in the tripped condition; otherwise, the
associated emergency diesel generator would be declared inoperable and
its applicable action statement(s) entered. Corresponding changes have
also been made in the bases for TS 3/4.3.2 and the BVPS-2 TS Index
pages.
Date of issuance: May 27, 1998.
Effective date: Effective immediately, to be implemented within 60
days (both units).
Amendment Nos.: 214 and 91.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19969).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 27, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, (BVPS-1 and BVPS-2)
Shippingport, Pennsylvania
Date of application for amendments: March 16, 1998, as supplemented
May 14, 1998.
Brief description of amendments: These amendments revise technical
specification (TS) Table 4.3-1 to add footnote 6 to the channel
calibration requirement for all instrument channels that are provided
with an input from neutron flux detectors. Footnote 6 provides that
neutron detectors may be excluded from channel calibrations. In
addition, BVPS-1 TS Table 4.3-1 is being revised to add channel
calibration requirements to items 2.b. (Power Range, Neutron Flux, Low
Setpoint), 5. (Intermediate Range, Neutron Flux), 6. (Source Range,
Neutron Flux (Below P-10)), and 23. (Reactor Trip System Interlocks P-
6, P-8, P-9, and P-10). Furthermore, changes are being made to correct
page numbers in the BVPS-2 TS Index and to add corresponding changes to
the TS Bases for both units.
Date of issuance: May 28, 1998.
Effective date: Both units, effective immediately, to be
implemented within 60 days.
Amendment Nos.: 215 and 92.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19969).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 28, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
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: January 9, 1998, as
supplemented by letter dated April 20, 1998.
Brief description of amendments: The amendments permit the use of
fuel with ZIRLO cladding.
Date of issuance: May 12, 1998.
Effective date: May 12, 1998.
Amendment Nos. 196 and 190.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9605).
The April 20, 1998 letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: December 10, 1997.
Brief description of amendment: The amendment clarifies sections of
the Technical Specifications that have been demonstrated to be unclear
or conflicting.
Date of Issuance: June 4, 1998.
Effective date: June 4, 1998, to be implemented within 30 days.
Amendment No.: 195.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4313).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated June 4, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station Unit No. 2, Oswego County, New York
Date of application for amendment: December 15, 1997, as
supplemented by letter dated April 24, 1998.
Brief description of amendment: This amendment changes Technical
Specifications 2.1.2 and 3.4.1.1 to revise the minimum critical power
ratio safety limits for fuel operating cycle 7 for two-loop and single-
loop recirculation operation.
Date of issuance: June 4, 1998.
Effective date: As of the date of issuance to be implemented before
[[Page 33117]]
startup of the Unit 2 reactor to begin fuel operating cycle 7.
Amendment No.: 82.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4314).
The April 24, 1998, submittal provided clarifying information that
did not alter the initial no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 4, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
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: September 2, 1997.
Brief description of amendment: The amendment corrects several
compliance issues as identified in Licensee Event Report 97-022-00
``Technical Specification Violations'' dated July 9, 1997, by rewording
the text; changing terminology and numbering; combining two Technical
Specifications (TSs) into one; changing the allowed outage times;
specifying guidance for entering into TS 3.0.3; changing a definition;
changing surveillance requirments, and updating the TS Bases section to
reflect changes.
Date of issuance: May 26, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 215.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50008).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 26, 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
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: October 15, 1997, as
supplemented January 23 and April 8, 1998.
Brief description of amendment: The amendment revises the action
statements and the instrumentation trip setpoint tables in the
Technical Specifications for the reactor trip system and engineered
safety feature actuation system instrumentation. In addition, the
amendment (1) decreases the reactor trip setpoint for the reactor
coolant pump low shaft speed (underspeed trip setpoint) from 95.8
percent to 92.4 percent of rated speed, (2) makes editorial changes,
and (3) changes the Bases to reflect the new methodology.
Date of issuance: May 26, 1998.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 159.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 19, 1997 (62
FR 61842).
The January 23 and April 8, 1998, submittals provided clarifying
and additional information that did not change the scope of the October
15, 1997, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 26, 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.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: April 7, 1998.
Brief description of amendment: The amendment replaces the
pressurizer maximum water inventory requirement with a pressurizer
maximum indicated level requirement. The amendment also makes editorial
changes and modifies the associated Bases section.
Date of issuance: May 27, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 160.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 23, 1998 (63 FR
20219).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 27, 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.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: April 14, 1998, as supplemented
May 7, 1998, and two letters dated June 4, 1998.
Brief description of amendment: The amendment changes Technical
Specification 3/4.4.4, Relief Valves, to ensure that the automatic
capability of the power-operated relief valves (PORVs) to relieve
pressure is maintained when these valves are isolated by closure of the
block valves. The amendment also makes editorial changes, adds PORV
surveillance requirements, and modifies the associated Bases section.
Date of issuance: June 5, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 161.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 20, 1998 (63 FR
19532).
The May 7, 1998, letter and the two letters dated June 4, 1998,
provide clarifying information that did not change the scope of the
April 14, 1998, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 5, 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
[[Page 33118]]
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: December 23, 1997.
Brief description of amendments: The amendments changed the
combined Technical Specifications (TS) for the Diablo Canyon Power
Plant (DCPP) Unit Nos. 1 and 2 to revise TS 3/4.7.1.1, Table 3.7-1,
``Maximum Allowable Power Range Neutron Flux High Setpoint With
Inoperable Steam Line Safety Valves.'' The power range (PR) neutron
flux high setpoints were changed based on revised calculational
methodologies for 1, 2, or 3 inoperable MSSVs per steam generator (SG).
The proposed TS change lowered the PR neutron flux high setpoints when
2 or 3 MSSV are inoperable per loop such that the maximum power level
allowed would be within the heat removing capability of the remaining
operable MSSVs. Although the method for calculating the maximum power
level allowed when one MSSV per loop is inoperable was revised, the
results were not and the limit remained the same. The associated Bases
were also revised.
Date of issuance: May 28, 1998.
Effective date: May 28, 1998, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1-125; Unit 2-123.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19975).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 28, 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.
Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear
Generating Station, Unit No. 1, Salem County, New Jersey
Date of application for amendment: March 26, 1998.
Brief description of amendment: The amendment revises Technical
Specification 3.1.3.3, ``Rod Drop Time,'' to change the applicability
from Mode 3 (hot shutdown) to Modes 1 and 2 (startup and power
operation).
Date of issuance: June 4, 1998.
Effective date: As of date of issuance to be implemented within 60
days.
Amendment No.: 211.
Facility Operating License No. DPR-70: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19978). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 4, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: May 30, 1997, as supplemented
April 1, 1998.
Brief description of amendments: The amendments revise the
Technical Specification requirements to reflect a design modification
that changes the power sources to valves associated with the low
pressure coolant injection mode of the residual heat removal system.
Date of issuance: June 2, 1998.
Effective date: As of the date of issuance to be implemented prior
to startup from the next refueling outage for both units.
Amendment Nos.: Unit 1-211; Unit 2-152.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 16, 1997 (62 FR
38139).
The April 1, 1998, submittal provided clarifying information that
did not change the scope of the May 30, 1997, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Southern Nuclear Power Company, Inc., Georgia Power Company, Oglethorpe
Power Corporation, Municipal Electric Authority of Georgia, City of
Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric
Generating Plant, Units 1 and 2, Burke County, Georgia
Date of application for amendments: November 20, 1997, as
supplemented by letter dated April 16, 1998.
Brief description of amendments: The proposed changes to the
Technical Specifications (TS): (1) Remove the inequalities applied to
the ``Trip Setpoint'' column of TS Table 3.3.1-1, ``Reactor Trip System
Instrumentation'' and TS Table 3.3.2-1, ``Engineered Safety Feature
Actuation System Instrumentation'' and revise the ``Trip Setpoint''
column to read ``Nominal Trip Setpoint;'' (2) Add footnotes (n) and (i)
to TS Tables 3.3.1-1 and 3.3.2-1, respectively, to include criteria for
channel operability, reset, and calibration tolerance about the trip
setpoint. These footnotes also allow for the trip setpoint to be set
more conservatively than the Nominal Trip Setpoint value as necessary
in response to plant conditions; (3) The Allowable Value for TS Table
3.3.1-1, Function 14.b, Turbine Trip--Turbine Stop Valve Closure, would
be revised from ``[greater than or equal to] 96.7% open'' to ``[greater
than or equal to] 90% open;'' (4) Revise footnotes (l) and (m) of TS
Table 3.3.1-1 to refer to Nominal Trip Setpoint and delete the
inequalities applied to the trip setpoints; (5) Delete the superscript
``(a)'' from the ``Trip Setpoint'' column on page 6 of 8 of Table
3.3.1-1; (6) Revise the inequality for the Engineered Safety Feature
Actuation System Allowable Value for Steam Line Pressure--Low (Table
3.3.2-1, Function 1.e) from ``[less than or equal to]'' to ``[greater
than or equal to];'' and (7) Revise associated TS Bases to reflect the
TS revisions.
Date of issuance: June 1, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1-101; Unit 2-79.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68318).
The supplement dated April 16, 1998, provided clarifying
information that did not change the scope of the November 20, 1997,
application and the initial proposed no significant hazards
determination.
[[Page 33119]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 1, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: May 1, 1995 (TXX-95090).
Brief description of amendments: These amendments revise section 3/
4.8.1 of the Technical Specifications (TSs) to reduce the minimum fuel
oil volume requirement during MODES 5 and 6 for an operable emergency
diesel generator (EDG) and allow continued OPERABLE status of diesel
generators during all MODES for 48 hours with greater than a 6 day
supply of diesel fuel for a given EDG.
Date of issuance: May 22, 1998.
Effective date: May 22, 1998, to be implemented within 30 days.
Amendment Nos.: Unit 1--Amendment No. 60; Unit 2--Amendment No. 46.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32373).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 22, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: December 4, 1997, as
supplemented by letters dated January 28, 1998, March 3, 1998, March 9,
1998, and April 24, 1998.
Brief description of amendment: The amendment permits the continued
used of the existing Siemens Power Corporation minimum critical power
ratio (MCPR) safety limits for WNP-2 Fuel Cycle 14 and changes the ASEA
Brown Boveri (ABB) MCPR safety limit for single loop operation from
1.08 for Cycle 13 to 1.09 for Cycle 14.
Date of issuance: May 29, 1998.
Effective date: May 29, 1998, to be implemented within 30 days from
the date of issuance.
Amendment No.: 154.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 14, 1998 (63 FR
2284).
The January 28, 1998, March 3, 1998, March 9, 1998, and April 24,
1998, supplemental letters provided additional clarifying information
and did not change the original no significant hazards consideration.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated May 29, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: February 25, 1998.
Brief description of amendment: The amendment revises the Technical
Specifications to implement performance-based containment leakage
testing under Option B of 10 CFR 50, Appendix J.
Date of issuance: May 28, 1998.
Effective date: May 28, 1998.
Amendment No.: 136.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 8, 1998 (63 FR
17237).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 28, 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.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: May 8, 1998, as supplemented by letter
dated May 11, 1998.
Brief description of amendment: The amendment adds a new Action
Statement to Technical Specification 3/4.3.2, Table 3.3-3, Functional
Unit 7.b., Refueling Water Storage Tank Level--Low-Low Coincident With
Safety Injection.
Date of issuance: May 28, 1998.
Effective date: May 28, 1998.
Amendment No.: 117.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (63 FR 26829 dated May 14, 1998). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by June 15, 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 amendment. The
Commission's related evaluation of the amendment, finding of exigent
circumstances, consultation with the State of Kansas and final
determination of no significant hazards consideration are contained in
a Safety Evaluation dated May 28, 1998.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for Licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Dated at Rockville, Maryland, this 10th 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. 98-16012 Filed 6-16-98; 8:45 am]
BILLING CODE 7590-01-P